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No. of Centre 8 2 4 5
9 Tc-99m In-111 Ga-67 Cr-51
8
7
6
5
4
3
2
1
0
Figure 13 shows the demand in PET radioisotope and radiopharmaceutical among centers in Malaysia. There
are 4 Fluorine-18 Based radiopharmaceuticals in demand which is Fluorine-18 FDG, Fluorine-18 FDOPA,
Fluorine-18 FCholine and Fluorine-18 Sodium Fluoride. While 2 Carbon-11 based radiopharmaceuticals which
isCarbon-11 Choline and Carbon-11 Acetate. The other PET radiopharmaceuticals that are in demand is
Galliium-68 Generator, Zirconium-89 Labeled Antibody and Copper-64.
4
3
3
2
No. of Centre 2
11111 11
1
0
Figure 1
Figure 14 shows the demand in types of radioiodine among centres. Iodine-131 and Iodine-123 shows high
demand where 8 centers require both of the radioiodine. While 5 centers demand for Iodine-123 Meta-
Iodobenzylguanidine. There other types of radioiodine only 1 center in demand of it which is Iodine-124, Iodine-
123 Orto Iodo-hippurate and Iodine-123 Iodobenzofuran.
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98 8
8
7
6 5
No. of Centre 5
4
3
2 11 1
1
0
Figure 15 shows the list of therapeutic radiopharmaceutical in demand among centers in Malaysia. The lists of
therapeutic radiopharmaceutical in demand are Yterium-90 Zevalin, Lutetium-177 Octreotate, Yterium-90
Microsphere, Rhenium-186 for bone pain palliation, Irridium-192 for Brachytherapy, Samarium-153 for Bone
pain Palliation and Yterium-90 for Synovectomy.
5
4
4
333
3
No of Centre 2
2
11
1
0
ioisotope / radiopharmaceutical that is
Table 1 shows the average cost of some medical radioisotope and radiopharmaceutical.
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Table 1: Average Cost (RM/unit) for Medical Radioisotope in Malaysia
Radioisotope Cost (RM)
Technetium-99m Generator RM 7500 /each
Gallium-67 RM 4000 / 5 mCi
Chromium-51 RM 3000 / mCi
Fluorine-18 FDG RM 15 / mCi
Gallium-68 Generator RM 70 000 12 0000 / each
Iodine-131 MIBG RM 3500/ mCi
Iodine-131 RM 20/ mCi
Yterium-90 Microspheres RM 24 000 / patient
DISCUSSION
38% respondents in this survey are from the Ministry of Health Hospitals with a Nuclear Medicine Services.
Currently, under the Ministry of Health, there are only 6 government hospitals that provide Nuclear Medicine
Services that are Kuala Lumpur Hospital, National Cancer Institute Putrajaya, Penang General Hospital,
Sultanah Aminah Hospital Johor Bahru, Sarawak General Hospital and Likas Hospital (Hospital Wanita dan
Kanak-Kanak, Sabah). Others respondent were from the Private Hospitals or Institutions and University
Hospitals. Most of the expertise groups that involve in this survey are from the Nuclear Medicine Physician /
Specialist and Medical officer group which is 52%, while 31% of them are from the pharmacist group. The
pharmacist was specifically chosen from those that working in the Nuclear Medicine Department or involve at
least in the procurement, preparation, quality control and dispensing of medical radioisotope &
radiopharmaceuticals to patient. In other words, they are best known as a Radiopharmacist or a Nuclear
Pharmacist.
The imaging modalities among these centers was assessed and most of the centers having either gamma camera
or SPECT-CT or both, as their imaging modalities. Only a few centers own a PET-CT Scan. Currently under
the Ministry of Health, there are only two centers having a PET-CT Scan that are National Cancer Institute
Putrajaya and Penang Hospitals. While the remaining are from the private and university hospitals.
In general nuclear medicine services, the most common medical radioisotope use is the Technetium-99m.
Technetium-99m is the backbone of nuclear medicine where almost all procedure in Nuclear Medicine involves
Technetium-99m such as the bone scan, brain scan, renal scan, and heart scan.
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Two centers utilize Chromium-51 and mainly used to measure the glomerular filtration rate (GFR) to assess
the renal function. Where else, only one center utilizes Gallium-67 and it is commonly used in the infection and
inflammation imaging.
The average cost of Technetium-99m generator ordered on weekly basis was currently RM 52 500 for a
generator capacity of 500 mCi 1000 mCi among the 7 centers while only 1 center used Technetium-99m
generator with an activity of more than 1000 mCi per week which amount to RM 7 500.
The highest demand of medical radioisotope in General Nuclear Medicine Services is Technetium-99m. In this
survey, 69% of Nuclear Medicine Centers utilizes Gamma Camera or SPECT for diagnostic procedure and the
radioisotope mainly used with the Gamma Camera or SPECT are Technetium-99m. As been mentioned before,
Technetium-99m is an established radioisotope with multiple uses and is a backbone of a Nuclear Medicine
Services. Currently the Technetium-99m Generator is being imported. But the Technetium-99m supply chain is
unstable globally. Lately, there are several forums and seminars held internationally addressing the issue on
the market supply of the Molybdenum-99 which is the parent of Technetium-99m. There are two main reactors
(Canada & Holland) that produces 60% of world Molybdenum-99 and in upcoming years, both reactors will be
shut down and supply of Mo-99 will be disrupted. This is why several countries look for an alternative way of
producing Technetium-99m. Among them is Canada which they took a lead in producing the Technetium-99m
using a Cyclotron instead of a Nuclear Reactor. The cyclotron produced Technetium-99m uses Molybdenum-
100 as a target material while reactor produced uses Molybdenum-99. The monograph of cyclotron produced
Technetium-99m is still not yet available.
Other medical radioisotopes that are in demand are Indium-111 (In-111), Gallium-67 (Ga-67) and Chromium-
51 (Cr-51). Indium-111 is commonly used in Octreotide scan, to detect tumour with a somatostatin receptor,
although it still can be used in infection imaging but less likely. Chromium-51 and Gallium-67 is actually is in
higher demand as compared to the current usage; five centers for Chromium-51 and four centers for Gallium-
67. The main reason why the usage for Chromium-51 and Gallium-67 is low while the demand is there is
because the cost of these medical radioisotopes is quite expensive and is currently imported. The average cost
for Gallium-67 and Chromium-51 are RM 4000 / 5 mCi and RM 3000 /mCi each.
Currently in Malaysia, the Positron Emission Tomography (PET) radioisotope and radiopharmaceutical that
are currently in used are Fluorine-18 Deoxyglucose (F-18 FDG) and Gallium-68 generator. F-18 FDG is the
most use PET radiopharmaceutical in diagnosis of cancer, inflammation and neurodegenerative brain disease.
F-18 FDG is produced by a cyclotron and has a short half-life of 109.7 minutes. Currently there are 4
cyclotrons operating in Malaysia. The details as in table 2. Cyclotron and PET Radiopharmaceutical
Preparation Facility need to be Good Manufacturing Practice (GMP) certified and the product manufactured
need to be registered with the Ministry of Health before it can be marketed out. Unlike the general nuclear
medicine services, such as the Technetium-99m, Chromium-51 and Indium-111, which only involve with the
preparation of radiopharmaceuticals, PET Cyclotron center need to comply with the cGMP requirement as it
involve with the manufacturing of radiopharmaceuticals. Moreover, the classifications of radiopharmaceuticals
for diagnosis are being move from Over the Counter (OTC) category to Poison category. Thus the registration
of the radiopharmaceuticals product is more stringent in to meet the requirement of the regulatory body;
National Pharmaceutical Control Bureau (NPCB), Ministry of Health.
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Table 2: Types of Cyclotron Available in Malaysia
Facility Model Manufacture Beam Energy Radioisotope
r Type Produced
National Cancer GE GE Healthcare H- / d - 16.5 MeV (H-) F-18 FDG
Institute, PETtrace
Putrajaya 8.5 MeV (d-)
BioMolecular Cyclone IBA H- / d - 18 MeV (H-) F-18 FDG
Industry
9 MeV (d-)
Beacon RDS 111 Siemens / CTI H- 11 MeV (H-) F-18 FDG
International
Medical Centre
Austral-Euro 370V Sumitomo H- / d - 18 MeV (H-) F-18 FDG
Diagnostic
9.5 MeV (d-)
Beside PET Radioisotope F-18 FDG, only one center utilizes Gallium-68. Gallium-68 is used in labeling with
peptides either DOTA-(Tyr3)-octreotate (DOTA-TATE) or DOTA-(Tyr3)-octreotide (DOTA-TOC), where
both are used in the diagnosis of neuroendocrine tumor. Gallium- to produce it as it
is eluted from a 68Germanium/68Gallium generator. Gallium-68 has a half-life of 68 minutes and the parent
radionuclide is Germanium-68 which has a half-life of 271 days. The generator can last up to a year where the
elution of Gallium-68 can be done every 4-5 hours after each elution. For each elution of Gallium-68 from the
generator, is enough for 3 to 5 patients to perform DOTA-TOC or DOTA-TATE PET/CT scan, depending on
the generator activity. The labeling of Gallium-68 with peptides is done within several minutes and it can be
done either manually or automated but radiation safety issue and the product sterility must not be
compromised. Commonly, Gallium-68 radiolabelling is done under laminar air flow which provides a clean
environment with final filtration to ensure the sterility of the product is intact. Preparation and radiolabeling
of Gallium-
registered as it is use in-house. Due to the short half-life of Gallium-68 (68 minutes), it is impossible for
Gallium-68 labeled peptides to be commercial out. However, the preparation of Gallium-68 in a Nuclear
Medicine Facility is still need to comply with the Good Preparation Practice (GPP) requirement as the product
is intended for human use.
The average price of F-18 FDG spend was calculated per order basis. For two PET/CT centers, the average
price was RM 6 000 with an average activity usage of around 100-200 mCi per day. Meanwhile, one PET/CT
center utilizes F-18 FDG with an activity between 200-500 mCi per day and the average cost is around RM 7
500 per day. Commonly, the PET/CT scan does not operate every day. Some PET/CT center does the scan 3
days in a week while some does less depend on each center. The average price for F-18 FDG spends was
calculated based on three PET/CT centers and there are still remaining PET/CT center that did not include
in this survey due to the non-responsive. The average cost of 68Germanium/68Gallium generator is in a range of
RM 70 000 to RM 120 000 each depends on the generator activity and the manufacturer.
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As for the PET medical radioisotope / radiopharmaceutical demand, Fluorine-18 based radiopharmaceutical is
still in demand. Besides F-18 FDG, the demand of F-18 based radiopharmaceuticals is Fluorine-18 FDopa for
brain imaging, F-18 FCholine for prostate cancer imaging and F-18 Sodium Fluoride for bone imaging. While
Carbon-11 based radiopharmaceuticals that are in demand are Carbon-11 Choline and Carbon-11 Acetate both
for prostate cancer imaging. However, currently in Malaysia, there is no cyclotron available with a Carbon-11
target. Most cyclotron center in Malaysia currently has Fluorine-18 and Nitrogen-13 target.
68Germanium/68Gallium generator is also among the PET radiopharmaceuticals that are in demand followed by
Zirconium-89 and Copper-64. Zirconium-89 has a half-life of 3.3 days and commonly Zirconium-89 is being used
to label it with monoclonal antibody for detecting various disease; commonly cancer.
As for radioiodine services, only two types of Iodine that are currently use that are Iodine-131 and Iodine-131
MIBG (meta-iodobenzylguanidine). In Malaysia, the highest usage is Iodine-131 as it is readily available with a
half-life of 8.02 days and it is currently used for both diagnosis and treatment of thyroid related disease. The
average utilization of Iodine-131 among 8 centers is 4200 mCi per week with an average cost of RM 84 000
while the utilization of Iodine-131 MIBG for 2 centers is 4.5 mCi per week with an average cost of RM 15 750.
Iodine-131 is mainly used as a therapy in the treatment of thyroid cancers, specifically papillary and follicular
thyroid cancer. In current practice, a low dose of Iodine-131 is also being used as for diagnostic purposes. It is
due to unavailability of diagnostic radioiodine; Iodine-123; in the local market. Iodine-131 labeled with MIBG
or also known as I-131-MIBG therapy is used in the treatment of tumors that specifically take up MIBG such
as neuroblastoma and pheochromacytoma.
The limitation of radioiodine isotope usage depends on mainly two factors that are:
i. Availability of Radioiodine isotope
Not all radioiodine isotopes are readily available in Malaysia such as Iodine-123 and Iodine-124.
Currently, the isotope of iodine that is being used is iodine-131.
ii. Radioiodine Preparation Facility
Preparation facility for oral radioiodine is only required if liquid dosage forms are being used.
Preparation facility is not needed for radioiodine capsule. The parameter to be considered when
handling liquid radioiodine is the dispensing area. At least two rooms are required for dispensing of
radioiodine to outpatient to ensure there is always a room available if spillage (e.g.; patient vomiting)
occurred in the other room. Centralized radioiodine preparation area and dedicated air conditioning
shall be considered. The preparation of radioiodine shall be carried out in a negative pressure control
area with a positive pressure airlock. At least, the radioiodine facility shall have dedicated rooms for
Radioiodine Preparation Room, Personnel Changing Room / Positive Airlock, Dispensing Room and
Radioactive Waste Room / Area. Not all Nuclear Medicine centers in Malaysia have a proper
radioiodine preparation facility.
The highest demand for radioiodine isotopes are Iodine-131 and Iodine-123. Iodine-131 is to be used mainly for
treatment of thyroid disorder while iodine-123 is to be used for diagnostic purposes. Iodine-123 is not available
in Malaysia due to its relatively short half-life of 13.22 hours and no neighboring countries are capable of
producing Iodine-123. The nearest country producing Iodine-123 are Korea and Japan. Other demand is Iodine-
124 which is to be used with a PET CT Scan. Iodine-124 has a half-life of 4.18 days and it is mainly for
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diagnostic purposes. The usage of PET-CT camera for radioiodine cases is interesting as the image resolution is
higher than the ordinary Gamma Camera or SPECT. Other radioiodine carriers that are in demand are for
Iodine-123 Meta-Iodobenzylguanidine (MIBG) which is use in diagnosis for the location of tumors such as
phaeochromocytomas and neuroblastomas, Iodine-123 Orto Iodo-hippurate to be used in renography and
Iodine-123 Iodobenzofuran (IBF) which is to be used in imaging of dopamine receptors.
Among the therapeutic radioisotope or radiopharmaceuticals that is currently being used among Nuclear
Medicine Centers in Malaysia are Yterium-90 (Y-90) Microspheres, Yterium-90 (Y-90) Synovectomy, Rhenium-
186 (Re-186), Irridium-192 (Ir-192) and Samarium-153 (Sm-153).
Yterium-90 Microsphere is used in the treatment of hepatocellular carcinoma or carcinoma of the liver. The
embolization of the microsphere within the cancer cells and the compound will then emit the radiation dose to
kill the cancer cell. There are two brand of Yterium-90 microspheres in the market that are SIR-Spheres (Y-90
resin microsphere) from Sirtex Medical limited and Therasphere from BTG International Ltd. The difference
between SIR-Spheres and Therasphere is that SIR-Spheres use biocompatible resin microspheres with size range
between 20 to 60 µm while Theraspehere use small glass microspheres with a size range between 20 to 30 µm.
Yterium-90 Synovectomy is being used mainly in the treatment of joint synovitis. Synovitis means
inflammation of the connective tissue lining a joint cavity or synovium. The radiopharmaceutical is
administered via joint puncture.
Samarium-153 is one of the radioisotopes that are approved in the USA and Europe for the palliation of pain
from metastatic bone cancer, whereas rhenium-186 is famously used by other countries.
A brachytherapy is known as a radioactive seeds that are placed in or near the tumor, which will then give a
high radiation dose to the tumor in order to kill the cancer cells. In other words it is in a form of a sealed
source. The radioisotope that is currently used in Malaysia for brachytherapy is the Iridium-192 in which its
decay is in β radiation.
The average numbers of patients undergone Yterium-90 Synovectomy is the highest among the therapeutic
radiopharmaceuticals in which is around 51 patients per year and followed by Yterium-90 Microsphere for
hepatocellular carcinoma which is around 24 patients per year. The usage of therapeutic radiopharmaceuticals
is not as high as the diagnostic radiopharmaceuticals are due to the cost. The cost to obtain therapeutic
radiopharmaceuticals is expensive and all are being imported. For example the average cost of Yterium-90
Microsphere therapy is around RM 24 000 per patient.
As for the demand, the highest are the Yterium-90 for Synovectomy, Yterium-90 Zevalin, Lutetium-177
Octreotate and Yterium-90 Microsphere. Yterium-90 Zevalin or Yterium-90 Ibritumomab tiuxetan is to be used
in the treatment of non- -177 Octreotate to be used in the
treatment of neuroendocrine tumour. There are two types of Lutetium-177 that are available worldwide, one is
carrier added and the other is the non-carrier added. The non-carrier added Lutetium-177 is currently being
favor as for its high specific activity (up to 3000 GBq/mg), high radionuclide purity and most important is
there is no long-lived isomer of Lutetium-177m despite its expensive cost.
CONCLUSION
The highest demand and the highest usage among all radioisotopes are Technetium-99m and Radioiodine
isotopes such as the Iodine-131, Iodine-131 MIBG, Iodine-123 and Iodine-123 MIBG. Currently, most of the
medical isotopes and radiopharmaceuticals are currently imported. Technetium-99m is the backbone of nuclear
medicine whereby more than 80% of Nuclear Medicine services utilize this radioisotope. Technetium-99m
supply chain is unstable globally and in coming future, two main reactors (Canada & Holland) that produces
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60% of world Molybdenum-99 will shut down the operation and supply of Molybdenum-99 will be disrupted
(OECD 2011). As for radioiodine services, currently, Iodine-
countries due to its short half-life. Iodine-123 is useful in diagnostic of thyroid related diseases. As for PET
services, the highest demands are F-18 FDG and Gallium-68 Generator for the moment. However the survey
data still did not include most of the PET centers in the Klang Valley, northern area (Penang) and the new
upcoming PET center in Southern Region (Malacca and Johor). It is important for Malaysia to self-produced
medical radioisotope and radiopharmaceuticals to meet the market and local demand of these medical isotopes.
More importantly, the emerging of new centers that offers nuclear medicine services and also the expansion of
the services of the existing nuclear medicine centers shall be taken into account.
ACKNOWLEDGEMENT
The authors wish to thank Dato Dr. Rehir bin Dahalan for his support of this work and Malaysian Nuclear
Agency, MOSTI.
REFERENCES
Kahn, Laura H.; The potential dangers in medical isotope production. Bulletin of the Atomic Scientists, 2008:
http://www.isotopeworld.com/filestore/Danger%20Medical%20Isotope%20pdf.pdf
naweb.iaea.org/napc/physics/research_reactors/database/RR%20Data%20Base/datasets/cateory/
status_operational_reactors.htm
OECD 2011, Nuclear Energy Agency organization For Economic Co-Operation and Development, The Supply
of Medical Radioisotopes: An Assessment of Long-term Global Demand for Technetium-99m
R.J. Kowalsky and S.W. Falen, 2013 ,Radiopharmaceuticals in Nuclear Pharmacy and Nuclear Medicine J Nucl
Med February 1 vol. 54 no. 2 324-325
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1Medical Technology Division, Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor, Malaysia.
2Department of Nutrition and Dietetics, Faculty of Medicine and Health Sciences, Universiti Putra
Malaysia, 43400 Serdang, Selangor, Malaysia.
3Department of Microbiology, Faculty of Biotechnology and Biomolecular Sciences, Universiti Putra
Malaysia, 43400 Serdang, Selangor, Malaysia.
* Corresponding author e-mail: [email protected]
ABSTRACT
Ficus deltoidea or locally known as Mas cotek is one of the common medicinal plants used in
Malaysia. Our previous studies showed that this plant have blood glucose lowering effect. Glucose
uptake into muscle and adipocytes cells is one of the known mechanisms of blood glucose lowering
effect. This study was performed to evaluate the effect of Ficus deltoidea on glucose uptake activity
into muscle cells. The cells were incubated with Ficus deltoidea extracts either alone or combination
with insulin. Amount of glucose uptake by L6 myotubes was determined using glucose tracer, 2-deoxy-
[1-3H]-glucose. The results showed that Ficus deltoidea extracts at particular doses enhanced basal or
insulin-mediated glucose uptake into muscle cells significantly. Hot aqueous extract enhanced glucose
uptake at the low concentration (10 µg/ml) whereas methanolic extract enhanced basal glucose uptake
at high concentrations (500 and 1000 µg/ml). Meanwhile, ethanolic extract enhanced glucose uptake at
low and high concentrations. Methanolic extract also mimicked insulin activity during enhancing
glucose uptake into L6 muscle cells. Glucose uptake activity of Ficus deltoidea could be attributed by
the phenolic compounds presence in the plant. This study had shown that Ficus deltoidea has the
ability to enhance glucose uptake into muscle cells which is partly contributed the antidiabetic activity
of this plant.
ABSTRAK
Ficus deltoidea atau tempatan yang dikenali sebagai Mas cotek merupakan salah satu tumbuhan
ubatan yang biasa digunakan di Malaysia. Kajian ini dijalankan sebelumnya menunjukkan bahawa
tumbuhan ini mempunyai kesan merendahkan glukosa darah. Penyerapan glukosa ke dalam sel-sel otot
dan adipocytes merupakan salah satu mekanisma yang dikenali glukosa darah mengurangkan kesan.
Kajian ini telah dilaksanakan untuk menilai kesan Ficus deltoidea aktiviti penyerapan glukosa ke
dalam sel-sel otot. Sel-sel yang telah incubated dengan Ficus deltoidea ekstrak sama ada bersendirian
atau kombinasi dengan insulin. Jumlah penyerapan glukosa oleh L6 myotubes adalah ditentukan
dengan menggunakan pengesanan glukosa, 2 - deoxy-[1-3H]-glukosa. Hasil kajian menunjukkan bahawa
ekstrak Ficus deltoidea pada dos tertentu dipertingkatkan penyakit atau sel-sel penyerapan glukosa
diantarai insulin ke dalam otot dengan ketara. Ekstrak air panas meningkatkan penyerapan glukosa
pada kepekatan rendah (10 µg/ml) manakala ekstrak methanolic meningkatkan penyerapan glukosa
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penyakit pada kepekatan yang tinggi (500 dan 1000 µg/ml). Sementara itu, ekstrak ethanolic
meningkatkan penyerapan glukosa rendah.
Keyword: Ficus deltoidea; glucose uptake; L6 myotubes; glucose tracer; phenolic compounds
INTRODUCTION
Diabetes mellitus is a metabolic disease characterized by persistent hyperglycaemia resulting from defects of
insulin secretion or insulin action or combination of these two factors (Alberti and Zimmet, 1998). It is the first
leading causes of death in developed country and has been an epidemic in many developing countries including
Malaysia (Ooyub et al., 2004). Globally, the prevalence of diabetes was estimated to be 2.8% in the year of
2000 and expected to rise to 4.4% in 2030. In Malaysia, it has been increased with an estimate number of
diabetes patients was 6.3% in 1986 and 14.9% in 2006 (Zanariah et al., 2008).
In Malaysia, there were 3.2 million (16.6%) of diabetes cases among adult in Malaysia in 2014. This prevalence
is projected to rise to 4.2 million (21.6%) in the year of 2020 (Saari and Noraini, 2013). The rising trend in the
prevalence of diabetes could possibly be due to the growth of population, aging, urbanization, and changes in
dietary habits, obesity and sedentary lifestyle (Zanariah et al., 2008). Diabetes mellitus can be controlled with
the uses of oral antidiabetic drugs. Even though plenty of antidiabetic drugs available, the disease remains
major global health problems. This could possibly be due to the drawback of the drugs such as adverse effects
and lack of efficacy (Kirchheiner et al., 2005). Therefore, searching for new antidiabetic agents should be
continued.
Ficus deltoidea, from Moraceae family, is one of the common medicinal plants in Malaysia. Based on
ethnobotanical approaches, this plant has been claimed to have antidiabetic properties (Mat-Salleh and Latif,
2002). Previous studies showed that this plant possess anti-hyperglycemic property (Adam et al., 2007; Adam
et al., 2010b; Adam et al., 2011a). Following 15-days treatment, hot aqueous extract of F. deltoidea stimulated
insulin release and reduced fasting hyperglycemia (Adam et al., 2011b). Elucidation of antihyperglycemic
mechanisms demonstrated that this plant enhanced basal and insulin-stimulated glucose uptake into liver cells
(Adam et al., 2009) and reduced the rate of glucose absorption from small intestine by inhibiting intestine
sucrase activity (Adam et al., 2010a). The present study was performed to find other possible antidiabetic
mechanisms of F. deltoidea, if any, by evaluating the potential of this plant to enhance glucose uptake into
muscle cells. The effect of F. deltoidea on glucose uptake activity was evaluated either in the absence (basal) or
presence (insulin-mediated) of 100 nM insulin.
MATERIALS AND METHODS
L6 cell line was purchased from American Type Cell Culture (ECACC, Salisbury UK). All cell culture
supplements were purchased from Invitrogen, USA. Ethanol and methanol were purchased from J.T. Baker
Chemical. Sodium chloride (NaCl), potassium chloride (KCl), calcium chloride (CaCl2), potassium dihydrogen
phosphate (KH2PO4), magnesium sulphate (MgSO4), sodium hydrogen carbonate (NaHCO3), HEPES, sodium
deodecyl sulphate (SDS), 3-(4, 5-dimethylthiazol-2-yl)-2,5-diphenyltetrazolium bromide (MTT), bovine insulin,
ammonium hydroxide (NH4OH), dimethylsulphoxide (DMSO), gallic acid, quercetin, 1,1-diphenyl-2-
picrylhydrazyl (DPPH), Folin- 3), aluminum chloride
(AlCl3), metformin and D-glucose were purchased from Sigma Chemical Co. (St. Louise, USA). Ultima GoldTM
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No. 2, December 2015.
LLT was purchased from PerkinElmer (USA). 2-Deoxy-[1-3H]-glucose was purchased from GE Healthcare
(USA).
Plants of F. deltoidea were collected at Sungai Tengi Selatan, Selangor, Malaysia. The plants were identified by
a taxonomist from the Biodiversity Unit of the Institute of Bioscience, Universiti Putra Malaysia (SK1467/07).
The leaves of F. deltoidea were oven dried at 45C and ground to a fine powder. Hot aqueous extracts were
prepared by boiling the powdered sample in distilled water for 3 hours (100 g/L) by changing the water every
hour. The combined suspension was filtered using Whatman filter paper No. 54 and freeze dried to give the
powdered form. Ethanolic and methanolic extracts was prepared by soaking the powder in 95% ethanol and
methanol, respectively for 3 days (100 g/L) at room temperature by changing solvent daily. The combined
suspension was filtered using whatman filter paper No. 54 and evaporated to dryness under pressure at 30C to
give ethanolic and methanolic extracts. The yields of the extracts were 21 g, 17 g and 19 g for hot aqueous,
ethanolic and methanolic extract, respectively.
The concentrations of phenolic compounds in the extracts of F. deltoidea were measured according to the
Folin-Ciocalteu method as described elsewhere (Sim et al., 2010). Briefly, 100 µl of 1mg/ml of F. deltoidea
extracts was added to 4.5 ml distilled water and 100 µl of 2N Folin-ciocalteu reagent with shaking for 3
minutes. Then, 200 µl of 2% (w/v) of sodium carbonate solution was added to the mixture. The reaction
mixtures were incubated in dark at room temperature for 3 hours. The absorbance was measured at 760 nm
using EnSpire Multimode Plate Reader (PerkinElmer, USA). All extracts were assayed in triplicate. Gallic
acid (0-1500 µg/ml) was used for calibration. Total phenolic content in the extracts is expressed as mg gallic
acid equivalence (GAE) per g extract.
The concentrations of phenolic compounds in the extracts of F. deltoidea were measured according to Yang et
al. (2011). Briefly, 150µl of 1mg/ml of F. deltoidea extracts was added to 150 µl of 2% (w/v) of AlCl3 solution
in 96-wells plate. The mixture was incubated in dark at room temperature for 15 minutes. The absorbance was
measured at 435 nm using EnSpire Multimode Plate Reader (PerkinElmer, USA). All extracts were assayed
in triplicate. Quercetin (0-100µg/ml) was used for calibration. Total flavonoids content in the extracts is
expressed as mg quercetin equivalence (QE) per g extract.
foetal bovine serum (FBS) and 1% (v/v) antibiotic solution (10,000 units/ml penicillin and 10 mg/ml
streptomycin) at 37C humidified with 5% CO2. Differentiation into myotubes was induced by reducing the
FBS in the complete culture medium from 10% to 2%. Cells were maintained in this medium for 4 to 6 days
post-confluence. The extent of differentiation was established by observing multinucleation of cells. In the
present experiment, about 85% - 90% of the myoblasts was fused into myotubes (Anandharajan et al., 2005).
The cells were seeded at concentration of 1.5x104 cells/well onto a sterile 96-wells plate and incubated at 37C
overnight. Cells were further incubated for 72 hours at 37C in the absence or presence of F. deltoidea extracts
(10-1000 g/ml) and metformin (10-2000 M). Endpoint measurement of viable cells was done according to
widely established methods (Mosman, 1983; Carmichael et al. 1987). Following the required incubation period,
20 l of 5 mg/ml of 3-(4, 5-dimethylthiazol-2-yl)-2, 5-diphenyltetrazolium bromide (MTT) was added to each
well and incubated for 4 hours. Subsequently, the media from each well was then gently aspirated and 100 l of
dimethylsulphoxide (DMSO) was added to dissolve the formazan crystals. Plates were shaken for 5 seconds and
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absorbance was measured at 570 nm using Anthos microplate reader (Beckman Coulter, USA). The percentage
of cell viability was calculated using the following formula:
% of cell viability = Absorbance of samples X 100
Absorbance of control
Glucose uptake assay was done according to Liu et al. (2001) with some modifications. Briefly, confluent cells
were seeded at a concentration of 2x105 cells/well in a 12-wells plate and left overnight at 37C to allow
attachment prior to test. The next day, cells were washed thrice with serum-free DMEM and pre-incubated
with this medium for 5 hours at 37 -Ringer
bicarbonate buffer (KRB). Cells were further incubated for 30 minutes at 37C with various concentrations of
F. deltoidea extracts (10-1000 g/ml) either alone or in combination with 100 nM insulin. Metformin was used
as positive control. To initiate glucose uptake reaction, 2-deoxy-[1-3H]-glucose (1 Ci/ml) diluted in 0.1 mM D-
glucose solution was added to each well and incubated further for 60 minutes at 37C. After incubation, cells
were washed thrice with ice-cold KRB buffer and solubilized with 0.1% sodium deodecyl sulphate (SDS)
dissolved in phosphate buffer, pH 7.4. The content of each well was transferred into scintillation vials and 15
ml of scintillation cocktail, Ultima GoldTM LLT was added. The radioactivity incorporated into the cells was
measured using Liquid Scintillation Counter (Hewlett Packard, USA).
All results are expressed as mean standard deviation for a given number of observations. Statistical analyses
were done using GraphPad Prism version 3 software. Data were analyzed using one way ANOVA followed by
Tukey post hoc test. Significant level was set at p<0.05.
RESULTS
In L6 myotubes, a significant reduction of cell viability occurs in the presence of hot aqueous extract at
concentrations of 50 1000 µg/ml, ethanolic extract at concentrations of 50 1000 µg/ml, methanolic extract
at concentrations of 100 1000 µg/ml and metformin at concentrations of 2000 and 5000 µM (Table 1).
Reduction of L6 myotubes in the presence of ethanolic extract at concentrations of 500 and 1000 µg/ml was
more than 50%.
The total phenolic and flavonoids content in F. deltoidea extracts are shown in Table 2. Methanolic and hot
aqueous contain high amount of phenolic compounds, whereas the ethanolic contains considerable amount.
Flavonoids also presence in the F. deltoidea extracts, however the amount of such metabolite is low.
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Table 1: L6 myotubes viability in the presence of F. deltoidea and metformin
Cells viability (%)
Test 0 g/ml/ 10 50 g/ml/ 100 500 1000
agents 0 M g/ml/ 500 M g/ml/ g/ml/ g/ml/
100 M 1000 M 2000 M 5000 M
Hot
aqueous 100.00 ± 94.77 ± 83.49 ± 83.08 ± 75.67 ± 70.71 ±
extract 5.70 6.57 6.31 *** 2.99 *** 5.22 *** 7.95 ***
(16.51%) (16.92%) (24.33%) (29.29%)
Ethanolic 100.00 ± 101.95 ± 81.17 ± 77.51 ± 40.54 ± 46.17 ±
extract 2.14 2.04 9.83 *** 7.72 *** 1.55 *** 2.19 ***
(18.83%) (22.49%) (59.46%) (53.83%)
Methanolic 100.00 ± 99.31 ± 91.92 ± 84.41 ± 81.28 ± 52.77 ±
8.90 8.41 *** 10.31 ** 1.42 ***
extract 1.58 1.68 (15.59%) (18.72%) (47.23%)
Metformin 100.00 ± 94.57 ± 99.65 ± 85.21 ± 79.92 ± 63.55 ±
14.19 19.54 15.18 ** 6.51 ***
14.24 18.72
(20.08%) (36.45%)
Notes: Cells were incubated for 72-hour in the presence of various concentrations of F. deltoidea extracts (10-
1000µg/ml) and metformin (100 5000 µM). Values expressed as percentage of Means ± Standard Deviations
(n=8) of the cells viability from three independent assays. *p<0.05, **p<0.01 and ***p<0.001 compared with
control. Values in the bracket indicate percentage of cell viability reduction.
Table 2: Total content results on Ficus deltoidea (Mas cotek)
Extract Total phenolic content Total flavonoids content
(Quercetin equivalence,
(Gallic acid equivalence,
g/mg extract) g/mg extract)
Methanolic 159.58 ± 6.89 19.52 ± 0.34
Ethanolic 49.58 ± 3.96 16.58 ± 0.50
Hot aqueous 126.67 ± 3.98 9.08 ± 0.37
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Percentage of 200 **
2-deoxy[1-3H]glucose uptake
150 *** *
relative to control
(dpm/105 cells/60min) 100
50
0 Basal With 100 nM
Control 100nM insulin
Insulin
10 g/ml 50 g/ml 100 g/ml 500 g/ml 1000 g/ml
Figure 1: Glucose uptake activity of F. deltoidea hot aqueous extract
Note: Values represent the means ± S.D from three independent experiments with four replicates in each
experiment. *p<0.05,**p<0.01;***p<0.001 compared with control incubation.
Hot aqueous extract showed a concentration-dependent decrease of basal and insulin-mediated glucose uptake
into muscle cells. Extract at concentration of 10 g/ml significantly enhanced glucose uptake by 1.28-fold
(p<0.05) and 1.30 (p<0.01) under basal and insulin-mediated state, respectively relative to control. 100 nM
insulin enhanced glucose uptake by 1.39-fold (p<0.001) relative to control.
250
Percentage of 200 *** * *** *** ** *** *** *** *** ***
2-deoxy[1-3H]glucose uptake 150
relative to control 100
(dpm/105 cells/60min)
50
0 100nM Basal With 100 nM
Control Insulin 100 g/ml insulin
10 g/ml 50 g/ml 500 g/ml 1000 g/ml
Figure 2: Glucose uptake activity of F. deltoidea ethanolic extract
Note: Values represent the means ± S.D from three independent experiments with four replicates in each
experiment. *p<0.05;***p<0.001 compared with control incubation.
Ethanolic extract significantly enhanced basal glucose uptake by 1.26-fold (p<0.05), 1.46-fold (p<0.001), 1.38-
fold (p<0.001) and 1.35-fold (p<0.001) at concentrations of 50, 100, 500 and 1000 g/ml respectively, relative
to control. Insulin-mediated glucose uptake was enhanced significantly at all concentrations evaluated with the
magnitude of uptake were 1.80-fold (p<0.001), 1.63-fold (p<0.001), 1.59-fold (p<0.001), 1.49-fold (p<0.001)
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and 1.46-fold (p<0.001) at concentrations of 10, 50, 100, 500 and 1000 g/ml, respectively compared to
untreated control.
300
Percentage of 200 *** ** *** ***
2-deoxy[1-3H]glucose uptake
*** ***
relative to control
(dpm/105 cells/60min) 100 **
0 Basal With 100 nM
Control 100nM insulin
Insulin
10 g/ml 50 g/ml 100 g/ml 500 g/ml 1000 g/ml
Figure 3: Glucose uptake activity of F. deltoidea methanolic extract
Note: Values represent the means ± S.D from three independent experiments with four replicates in each experiment.
**p<0.01;***p<0.001 compared with control incubation. p<0.001 compared to 100 nM insulin.
Methanolic extract enhanced basal glucose uptake by L6 myotubes in a concentration-dependent manner.
Extract at concentrations of 500 and 1000 g/ml exhibited a significant enhancement of glucose uptake by
1.27-fold (p<0.01) and 1.86-fold (p<0.001), respectively relative to control. The enhancement by the later
concentration was 1.27-fold (p<0.001) higher than that of 100 nM insulin which evoked an uptake of 1.46-fold
(p<0.001) relative to control. Insulin-mediated glucose uptake was significantly enhanced by 1.27-fold
(p<0.001), 1.25-fold (p<0.01), 1.56-fold (p<0.001) and 1.33-fold (p<0.001) at concentrations of 10, 50, 100 and
500 g/ml respectively.
Percentage of 2-deoxy[1-3H]glucose 400
uptake relative to control 300
(dpm/105 cells/60min) *** ***
200 *** *** ***
100 ***
*** **
0 Insulin Basal With Insulin 100nM
Control 100nM
50M 100M 500M 1000M 2000M
Figure 4: Glucose uptake activity of metformin
Note: Values represent the means ± S.D from three independent experiments with four replicates in each experiment.
**p<0.01;***p<0.001 compared with control incubation. p<0.001 compared to 100 nM insulin.
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Basal glucose uptake following metformin treatment was enhanced significantly by 2.41-fold (p<0.001) and
2.76-fold (p<0.001) at concentrations of 1000 and 2000 µM, respectively compared to control. The enhancement
by these concentrations were 1.32-fold (p<0.001) and 1.51-fold (p<0.001) higher than that of 100 nM insulin
which evoked a 1.83-fold (p<0.001) of uptake relative to control. The later concentration exhibited the highest
basal glucose uptake and was used as positive control to challenge the effect of F. deltoidea extracts on basal
glucose uptake into L6 myotubes. Under insulin-mediated state, enhancement of glucose uptake were in the
range of 1.68 3.58-fold (p<0.01; p<0.001) compared to control and 1.63 1.95-fold (p<0.001) compared to
100 nM insulin. The highest insulin-mediated glucose uptake was found at concentration of 100 µM which
evoked a 3.58-fold of enhancement. Therefore, metformin at this concentration was used as positive control to
challenge insulin-mediated glucose uptake activity of F. deltoidea on L6 myotubes.
DISCUSSIONS
The present study reports the effect of hot aqueous, ethanolic and methanolic extracts of F. deltoidea on in
vitro basal and insulin-mediated glucose uptake activity into skeletal muscle using L6 myotubes as the model of
muscle cells. This cell has been widely used as the model system for studying glucose uptake activity into
skeletal muscle (Patel and Mishra, 2008). This study revealed that F. deltoidea extracts at particular
concentrations have the ability to enhance basal and insulin-mediated glucose uptake into L6 muscle cells. This
suggests that there is a possibility of presence of antidiabetic compounds in F. deltoidea extracts which exert its
antidiabetic mechanism through the enhancement of glucose disposal into muscle cells.
The viability of L6 myotubes in the presence of F. deltoidea extracts was evaluated using MTT assay. In this
assay, the yellow tetrazolium salt, MTT is reduced by the mitochondrial enzymes, succinate dehydrogenase to
form insoluble purple formazan crystals which are solubilized by the addition of a detergent. The color
produced then can be measured spectrophotometrically at 570 nm. MTT reduction was proportional to cell
viability (Mosman, 1983; Carmichael et al. 1987). In cytotoxicity evaluation, the highest concentration of a test
agent should be 1000 g/ml or 1000 M. If none of the concentrations of test agents exhibited cytotoxic effect
in excess of 50% of cell populations, the test agent is considered non-toxic against the tested cell line (Elmore et
al., 2002). In the present study, ethanolic extract at concentrations of 500 and 1000 µg/ml reduces the viability
L6 myotubes to less than 50% after 72 hours exposure. Thus, the ethanolic extract was considered toxic against
the L6 myotubes (Elmore et al., 2002) and glucose uptake activity possess by these concentration of ethanolic
extract should not be taken into account.
Glucose uptake by insulin-targeted cells (liver, adipocytes and muscle cells) becomes the initial step in the
process of glucose disposal from blood circulation (Olson, 2012). Insulin, an anabolic hormone secreted by
pancreatic -cells was found to mediate glucose uptake into liver, adipocytes and muscle cells by binding to
insulin receptor (IR) proteins at the surface of the cells, activating a series of proteins within the cells which
leads to the translocation of glucose transporter 4 (GLUT4) protein to cell surface and permit the entrance of
glucose into the cells (Watson and Pessin, 2006; Roffey et al., 2007). It was reported that insulin enhanced
basal glucose uptake into muscle cells (Mitsumoto et al., 1991). The present study was in agreement with such
reports that insulin at concentration of 100 nM significantly enhanced glucose uptake by 1.39 - 1.83-fold in L6
myotubes. Therefore, this concentration of insulin was used in this study to mediate glucose uptake activity
into muscle cells by F. deltoidea extracts. This concentration of insulin (100 nM) was also been used to mediate
glucose disposal into adipocytes cells (Konrad et al., 2002; Sakurai et al., 2004).
Basal glucose uptake was significantly enhanced in the presence of hot aqueous extract (10 g/ml), ethanolic
extract (50, 100, 500 and 1000 g/ml), methanolic extract (500 and 1000 g/ml) and metformin (1000 and
2000 M). The enhancement of basal glucose uptake activity by methanolic extract at concentration of 1000
g/ml and metformin at concentrations of 1000 and 2000 M were significantly higher than that of 100 nM
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insulin. These results indicate that methanolic extract and metformin possess insulin-mimetic activity during
enhancing glucose uptake into L6 myotubes cells. Several antidiabetic plants such as Lagerstreoma speciosa
Agaricus campestris and Trigonella foenum-graecum have been reported to possess insulin-mimetic activity
(Liu et al., 2001; Vijayakumar et al., 2005). Trigonella foenum-graecum was reported to mediate glucose uptake
into adipocytes and liver cells through the activation of tyrosine phosphorylation of β-subunit of insulin
receptor (IR), subsequently enhancing tyrosine phosphorylation of insulin receptor substrate-1 (IRS-1) and p85
subunit of phosphatidylinositol 3-kinase (PI3-Kinase) and lead to glucose uptake (Vijayakumar et al., 2005).
There is a possibility that F. deltoidea extracts also mimic insulin mimetic activity through the same
mechanisms. However, to confirm this suggestion, further evaluations are needed to be carried out.
The insulin-mimetic activity of some plants during enhancing glucose uptake into targeted cells has been
reported to be associated with the phenolics compounds in the plants. For example, gallic acid, a type of
phenolic acid was reported to increase GLUT4 translocation and induce glucose uptake into 3T3-L1 adipocytes
(Prasad et al., 2010). Patel et al., (2012) reported that insulin-mimetic activity of many antidiabetic plants
were attributed to the presence of phenolic, flavonoids, terpenoids, coumarin and other constituents. It was also
reported that procyanidins, a flavonoid from grape seed possess insulin-mimetic activity during stimulating
glucose uptake into 3T3-L1 adipocytes cells (Pinent et al., 2004). In this study, it was shown that methanolic
extract of F. deltoidea contains high amount of phenolic compounds. Methanolic extract also contain moderate
amount of flavonoids. Thus, there is possibility that insulin-mimetic activity showed by methanolic extract is
attributed by phenolic compounds, including flavonoids presence in the extract.
Insulin-mediated glucose uptake into L6 myotubes was significantly enhanced in the presence of hot aqueous
extract (10 g/ml), ethanolic extract (all concentrations), methanolic extract (10, 50, 100 and 500 g/ml) and
metformin (all concentrations). The enhancement of insulin-mediated glucose uptake into this cell by
metformin at concentrations of 50, 100 and 500 M was significantly higher than that of 100 nM insulin. This
result suggests that metformin possess insulin-sensitizing activity during enhancing insulin-mediated glucose
uptake into L6 myotubes (Ko et al., 2005; Benhaddou-Andaloussi et al., 2008). This was in accordance with the
previous studies which reported that metformin sensitize the action of insulin on glucose uptake into muscle
cells (Klip and Leiter, 1990). However, all F. deltoidea extracts did not exerted insulin-sensitizing activity
during enhancing glucose uptake into L6 muscle cells.
Metformin was used as positive control to challenge glucose uptake activity of F. deltoidea extracts into L6
myotubes. Metformin is an effective antihyperglycemic agent with the main antidiabetic mechanism is through
the enhancement of insulin-mediated glucose uptake in muscle tissue and reduction of gluconeogenesis in the
liver, resulted in reduction of hyperglycemia (Chehade and Mooradian, 2000). It was shown that, under both
basal and insulin-mediated states, the magnitude of glucose uptake by all F. deltoidea extracts were less than
metformin. This indicates that glucose uptake activity of F. deltoidea into L6 myotubes was not potent as
metformin. This could possibly be due to that the plant extracts consist of mixture of compounds which are
bioactive and non-bioactive. There is possibility that the non-bioactive compounds antagonized the action of
the active compounds, hence decrease the glucose uptake activity of the extracts. Unlike the extracts,
metformin is a single compound and its potential to stimulate glucose uptake into muscle cells has been
scientifically proven and documented (Viljanen et al., 2005; Amini et al., 2005). Therefore, to ensure the
effectiveness of F. deltoidea extracts in enhancing glucose uptake into the targeted cells, the bioactive
compounds should be isolated and purified.
CONCLUSION
The results showed that aqueous, ethanolic and methanolic extracts of F. deltoidea have the ability to enhance
basal and insulin-mediated glucose uptake into L6 muscles cells. The plant possesses insulin-mimetic activity
during enhancing glucose uptake into such cells. From this study, it is suggested that the antihyperglycemic
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action of F. deltoidea extracts is mediated partly, by enhancement of glucose uptake into the muscle cells. This
study provide pre-clinical pharmacological data on the anti-diabetic activity of F. deltoidea and this
information can be used to support the development of F. deltoidea -based phytomedicine for diabetes
management.
ACKNOWLEDGEMENTS
The authors wish to thank Malaysian Nuclear Agency for providing facilities for this research.
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1Medical Technology Division, Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor
2Nuclear Medicine Department, National Cancer Institute, 62250 Putrajaya
ABSTRACT
PET/CT Scan with 68Ga-labelled analogs is of increasing interest in Nuclear Medicine and currently
is being performed all over the world. However, such labeling procedure requires high purity and
concentrated solutions of 68Ga. Here we report the purification and concentration of the eluate of
SnO2-Based 68Ge/68Ga generators via the anion exchange method. Three different anion columns were
selected to purify and concentrate the68Ga eluates which are Chromafix, Oasis WAX and AG 1-X8
columns. The different anion columns were compared and evaluated in terms of their capability in
adsorption and desorption of 68Ga from the generator. While the optimum molarity of Hydrocholric
Acid (HCl) for highest 68Ga retention was also determined starting from the ranges of 4M to 7M of
HCl. The results showed that the percentage of 68Ga retention or adsorption started to be plateau at
molarity of 5.5M for all three anion cartridges. One-way ANOVA analysis proved that there is no
significant difference between 5.5M with 6.0M, 6.5M and 7.0M which means that the retention of
gallium-68 is equal at those molarities. At 5.5M, Chromafix and Oasis WAX cartridges showed the
highest retention of gallium-68 which is 98.30%. The lowest 68Ga retention was gained by AG 1-X8
column which is 97.07%. While for desorption of 68Ga, the highest percentage was obtained by using
Oasis WAX cartridges which is 70.49% followed by Chromafix which is 70.36%. The lowest desorption
of gallium-68 was obtained by using AG 1-X8 column which is only 58.56%. Therefore, from this
study, the most suitable cartridge and HCl molarity that should be applied in purification and
concentration of Gallium-68 eluate from a SnO2 Based 68Ge/68Ga generator is Oasis® WAX column
with a HCl molarity of 5.5M respectively.
abstracts margin ABSTRAK
not aligned
PET / CT Scan dengan analog 68Ga-labelled adalah faedah yang meningkat dalam Ubat Nuclear dan kini
sedang diusahakan seluruh dunia. Bagaimanapun , prosedur melabelkan sedemikian memerlukan kesucian yang
tinggi dan larutan pekat 68Ga. Di sini kami melaporkan penulenan dan tumpuan eluat 68Ge -Based SnO2 /
generator-generator 68Ga melalui kaedah pertukaran anion. Tiga tiang-tiang anion yang berbeza dipilih
menyucikan dan menumpukan the68Ga eluates yang mana ialah Chromafix, Oasis WAX and AG tiang-tiang 1-
X8. Tiang-tiang anion yang berbeza telah dipertandingkan dan dinilaikan dalam soal kemampuan mereka dalam
penjerapan dan desorption of 68Ga dari penjana. Manakala kemolaran optimum Hydrocholric Acid untuk
pengekalan 68Ga tertinggi juga ditentukan mula dari banjaran 4M kepada 7M of HCl. Keputusan menunjukkan
bahawa peratusan pengekalan 68Ga atau penjerapan dimulakan untuk menjadi dataran tinggi di kemolaran
5.5M untuk semua tiga sarung peluru anion. Analisis ANOVA yang sehala membuktikan bahawa tiada
perbezaan penting antara 5.5M dengan 6.0M, 6.5M and 7.0M yang bermaksud pengekalan galium 68 menyamai
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No. 2, December 2015.
di yang molarities. Di 5.5M, kartrij-kartrij Chromafix and Oasis WAX menunjukkan pengekalan tertinggi
galium 68 yang mana 98.30%. Pengekalan 68Ga terendah telah diperolehi oleh AG lajur 1-X8 yang mana
97.07%. Manakala untuk desorption of 68Ga, peratusan tertinggi telah diperolehi dengan menggunakan kartrij-
kartrij Oasis WAX yang mana 70.49% diikuti oleh Chromafix yang mana 70.36%. Penyaherapan terendah
galium 68 telah diperolehi dengan menggunakan AG lajur 1-X8 yang mana hanya 58.56%. Lantarannya,
daripada kajian ini, kartrij paling sesuai dan kemolaran HCl yang seharusnya digunakan dalam penulenan dan
tumpuan 68 eluat, Gallium dari SnO2 Based 68Ge / penjana 68Ga ialah lajur Oasis® WAX dengan kemolaran
HCl 5.5M masing-masing.
Keywords: Anion Exchange, Gallium-68, PET/CT Scan, Germanium-68, Generator
INTRODUCTION
Nuclear medicine is a specialty of medicine and medical imaging that uses radionuclides (radiopharmaceuticals)
in a very small amount and relies on the process of radioactive decay in the diagnosis and treatment of disease.
It is the safe use of radioactive materials in the diagnosis and treatment of various diseases. In imaging, the
radiopharmaceuticals are detected by special types of cameras like gamma camera that work with computers
that have special software in order to provide very precise pictures about the area of the body being imaged.
For a treatment purpose, the radiopharmaceuticals go directly to the organ being treated. The amount of
radiation in a nuclear typical treatment is kept within safe limits.
Certain compounds (pharmaceuticals) which by their nature, concentrate in different organs of their human
body, are chemically labelled with specific radioactive materials (radioisotopes). These radiopharmaceuticals,
once administered, concentrate within the organ or organ system and the distribution is determined by
specialised equipment.
Nuclear medicine was started with the invention of the cyclotron by Ernest Orlando Lawrence (1901-1958) and
the widespread clinical use of nuclear medicine began in the early 1950s. In Malaysia, nuclear medicine unit was
formed in 1964 as a subunit under the Department of Radiotherapy, Kuala Lumpur Hospital.
In nuclear medicine procedures, elemental radiopharmaceuticals are combined with other elements to form
chemical compounds, or else combined with existing pharmaceutical compounds, to form radiopharmaceuticals.
These radiopharmaceuticals, once administered to the patient, can localize to specific organs or cellular
receptors. This property of radiopharmaceuticals allows nuclear medicine the ability to image the extent of a
disease-process in the body according to the cellular function and physiology of particular organs.
Positron emission tomography (PET) is an imaging method in nuclear medicine. It combines the potential to
quantify the tracer uptake within lesions with a relatively high resolution. High resolution means that the
picture of a specific organ or area in the human body can be displayed in a high quality. In addition,
-emitting isotopes such as 11C, 13O, 15N which
render any molecule chemically unchanged compared to the original molecule.
The most commonly used PET-radionuclide is 18F and is being produced in a cyclotron. Another strategy to
produce positron-emitting radionuclides is via a generator such as 68Ge/68Ga, 82Sr/82Rb and the 62Zn/62Cu
generators. One of the advantages of generators is they allow clinical studies without an on-site cyclotron or if
cyclotron beam time may not be available.
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Gallium is the third element of Group 13 of the Periodic Table. The +3 oxidation state of this metal is the
most stable in aqueous solution. Ga2+ would exist in the presence of high concentration of Cl- ions but it is
thermodynamically unstable to further oxidation to Ga3+. In aqueous solution, the free hydrated Ga3+ ion is
stable only under acidic conditions. In the pH range of 3-7, it can hydrolyse to insoluble trihydroxide if its
concentration exceeds nanomolar level. However, this precipitation can be avoided in the presence of stabilizing
agents. At physiological pH, the solubility of gallium is high due to the almost exclusive formation of
[Ga(OH)4]- ions.
In order to be suitable as a radiopharmaceutical, a Ga3+ chelate has to be thermodynamically stable towards
hydrolysis at physiological pH or to be kinetically stable during the period of clinical use. Another requirement
is the Ga3+ chelate also does not undergo exchange with the abundant blood serum protein transferrin.
The 68Ge/68Ga generator provides an excellent source of positron-emitting 68Ga. Radionuclide generators have
the advantage of providing radionuclides on demand. This is potentially an inexpensive and convenient
alternative to the on-site cyclotron production of short-lived radionuclides. Radionuclide generators usually
contain a long-lived parent radionuclide which decays to a short-lived daughter. There needs to be a sufficient
chemical difference to allow efficient chemical separation of the stationary parent from the soluble daughter.
The development of the 68Ge/68Ga generator has been reviewed in several articles (Lambrecht & Sajjad 1988;
Mirzadeh & Lambrecht 1996). The long half-life (t1/2 =270.8 days) of the parent 68Ge combined with the half-
life of 68Ga (t1/2=68 min), suitable for radiopharmaceutical synthesis, makes this pair ideal for a generator
strategy.
The preferred production route of 68Ge is via the (p, 2n) reaction on gallium targets (Ga2O3). This reaction
provides a significant cross section but experimental yields amount to 0.74 MBq/μ A/h only; therefore high
current accelerators are needed for sufficient batch yields (Rösch and Knapp 2003).
An important aspect for wide use of 68Ga in clinical PET is its chemical form and concentration after elution
from the generator. In addition, there is concern about 68Ge-breakthrough and contamination of the generator
column material. Nowadays, TiO2-based generator has become commercially available.
Due to the long half-life of 68Ge, a good separation system of mother and daughter to avoid breakthrough is
mandatory. Two different strategies have been employed to afford this separation of 68Ge (IV) from 68Ga (III).
The presence of metal ions, especially Zn in the eluate is frequently a concern for labelling DOTA-peptides with
radiometals as 68Ga, since the incorporation of radiometals in DOTA-peptides is negatively influenced
(Breeman et al., 2005, 2003). Zn ions will always be present in the eluate of 68Ge/68Ga generators, due to
formation of 68Zn as decay product of 68Ga.
The amount of formed 68Zn is dependent on the time between prior elution. All other metals are low (<1 ppm),
and Ti) is low, their concentration can be much higher than the concentration of 68Ga.
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The most popular method for the purification of charged molecules is ion exchange chromatography which also
was known as ion exchange column. In anion exchange column, negatively charged molecules are attracted to a
positively charged solid support. Conversely, in cation exchange chromatography, positively charged molecules
are attracted to a negatively charged solid support. Anion exchange columns include AG 1-X8 resins,
Chromafix and Oasis Wax cartridges.
In order to optimize of all charged molecules, generally, the mobile phase is a low to medium conductivity
solution. The adsorption of the molecules to the solid support is driven by the ionic interaction between the
oppositely charged ionic groups in the sample molecule and in the functional ligand on the support. . The
strength of the interaction is determined by the number and location of the charges on the molecule and on the
functional group.
By increasing the salt concentration (generally by using a linear salt gradient) the molecules with the weakest
ionic interactions start to elute from the column first. Molecules that have a stronger ionic interaction require
a higher salt concentration and elute later in the gradient. The binding capacities of ion exchange resins are
generally quite high. This is of major importance in process scale chromatography.
Although the commercially available 68Ge/ 68Ga generator system provides an excellent source of 68Ga but
not optimized for the routine synthesis of 68Ga-labelled radiopharmaceuticals. The eluates have low specific
volume of 68Ga, high acid concentration and different trace elements. Hence, pre-concentration and purification
of the initial generator eluate should be done.
BACKGROUND
The 68Ge/68Ga radionuclide generator provides an excellent source of positron emitting 68Ga for the routine
synthesis and application of 68Ga-labelled peptides using positron emission tomography (PET). Currently, the
method of 68Ga purification from SnO2-based 68Ge/68Ga generator is via fractionation method. There are several
problem imposed by adopting the fractionation method in the radiolabelling processes.
One of them is low percentage of recovery of Ga-68. Furthermore, the current eluation system of the Ga-68
produces low specific volume of 68Ga eluate within the 2 ml fraction of 0.6M HCl. One of the factors that
have low
specific volume of 68Ga, high acid concentration, a breakthrough of 68Ge of 10-2% increasing with time or usage
frequency and impurities such as stable Zn(II) generated by the decay of 68Ga and Fe(III) as a general impurity
(Decristoforo et al., 2007; Erik et al., 2010; Mattia et al., 2008;). Hence, pre-concentration and purification of
the initial generator eluate is mandatory.
Secondly, the obtained volume of Ga-
remains within the outlet tubing between the generator and the reactor. This condition will cause the
adjustment of pH in the reactor become unpredictable and finally may lead to inconsistent labelling yield.
Thirdly, high acidity of Ga-68 eluate (0.6M in 2ml fraction) from the generator will affect the radio labelling
activity. The rather large volume (up to 10 ml for complete elution) and high proton concentration of the
generator eluate require pre-concentrationof the activity for labeling nanomolar amounts of peptides. For all
these reasons, purification and concentration of the 68Ga-eluate must be performed before labelling step (Mattia
et al., 2008).
Currently, the method that are being practiced for the SnO2 Based Generator is by fractionation method in
order to purify and concentrate the Ga-68 eluate. However, this kind of method does not totally eliminate the
metallic impurities instead of only decrease the impurities (Mattia et al., 2008). Purification method of the
generator-produced 68Ga and requires an eluate fractionation ( Breeman et al., 2005). Although this approach
does not eliminate but only decreases the metallic impurities, it has been successfully employed for 68Ga-
DOTATOC synthesis (Breeman et al., 2005; Decristoforo et al., 2007).
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One of the methods to purify Ga(III) is by using a cation-exchange column which eluted with HCl/acetone
mixtures was reported (Strelow et al., 1971). 68Ga in 0.1 M HCl can be quantitatively absorbed on acation-
exchange resin and further purified from Ge, Ti, Fe, Zn and Al impurities by adjusting the HCl concentration
and HCl/ acetone ratio of the eluent (Zhernosekov et al., 2007). Finally, 68Ga can be released from the resin in
a chemical form suitable for labelling by increasing the acetone content. The overall content of acetone in the
purified Ga-68 fraction is small and non-toxic. However, this method could not be done in the research project
because the 68Ge/ 68Ga generator that has been used in the experiments was using 0.6M HCl to eluate the Ga-
68 from the generator where the eluate is in Ga3+ ions form. Ga3+ ions have higher tendency to form the
negatively charged gallium tetrachloro complex [GaCl4]- complex that can be adsorbed on a strong anion
exchange resin (Meyer et al., 2004) instead of cation exchange column.
Purification and concentration procedures of Ga-68 eluate also can be done via solvent extraction method
where an organic solvent has been used like methyl ethyl ketone (MEK). According to T.H. Bokhari et al.
2008, this method showed that at 1:2 ratio of eluate and MEK, almost 100% Ga-68 activity is extracted in
MEK. The extraction is finished within 2 to 4 minutes. The mixing and shaking steps had been done for 4
minutes. Then, the organic solvent (MEK) was evaporated. The purpose of doing so is to concentrate the Ga-
68 eluate. The activity of 68Ga recovered was noted after dissolving it in buffer solution. Finally, the whole
procedure of 68Ga concentration was finished in 10 min. Although the yield or outcome of the solvent extraction
method is good but there are a lot of modifications must be done before running the procedures and it is a
time-consuming method. Therefore, this method has not been recommended for the research project.
Recently, another method to purify and concentrate Ga-68 eluate is by increasing the HCl concentration to ~6
M and passing the mixture through an anion exchange column where the anionic [GaCl6]3 and
[GaCl4] complexes are strongly absorbed, while 68Ge(IV), Al(III), Ti(IV) and In(III) are practically passed
through the column. 68Ga is then eluted with <200 μl of pure water [Breeman et al., 2005; Decristoforo et al.,
2007; Velikyan et al., 2004).
Furthermore, in this project, anion-exchange column will be used as our method in order to achieve objectives
of the project. Ga3+ ions form the negatively charged gallium tetrachloro complex [GaCl4]- at concentration of
HCl is more than 5.0 M. Therefore, this feature can be used advantageously to concentrate the eluted Ga-68,
since the [GaCl4]-complex can be adsorbed quantitatively on a strong anion exchange resin according to Meyer
et al. (2004).Ga-68 activity in 6 ml 5 M HCl was adsorbed (>98%) on all anion columns.
Desorption of Ga-68 activity from total Ga-68 activity on anion columns ranged between 50% and 83%. The
highest desorption was obtained with the 30mg weak anion Oasis WAX column where 83% was desorbed in
1mL Milli-Q. In order to reduce acidity of the eluate of anion column, H+ was replaced by Na+ whilst keeping
[Cl-] constant at 5 M. Furthermore, no statistically significant changes in Ga-68 desorption were observed while
decreasing at constant [Cl-] (Erik de Blois et al. 2010).
Thus, in this study, anion-exchange resin has been chosen rather than fractionation because it can be used to
reduce ionic impurities, to increase the concentration of generator eluate and to reduce acidity (Meyer et al.,
2004; Velikyan et al., 2004; Zhernosekov et al., 2007). Although fractionation results in a ready to use eluate
containing approximately 80% of the elutable Ga-68 activity but here we must be noticed that the major
limitations for direct use of Ga-68 for radiolabelling of peptide (DOTA-TATE) are the large volume of
generator eluate, high [H+], Ge-68 breakthrough and also potential metal ion impurities (Erik de Blois et al.
2010).
Furthermore, a study by Velikyan et al. (2004) describes most suitable volume reductions for the 68Ga eluate in
which they purified and concentrated the radionuclide by anion exchange resins. By this means the volume of
the final aqueous elute could be further decreased to 50-200 µl thus obtaining a high volume activity of 5 to 6
MBq/ µl using a 68Ge/68Ga generator of ~1250MBq initial 68Ga activity (Gebhardt et al., 2010).
Another reason that supports the method of choice is although in fractionation method, contents of 68Ge and
metallic impurities are lowered principally but not chemically removed (Zhernosekov et al. 2007).
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The paper describes the method to purify and concentrate Ga-68 eluate from a SnO2-based 68Ge/68Ga generator
by using anion-exchange purification method; compare the retention and desorption of Ga-68 eluate by using
three different types of anion cartridges; identify the optimum hydrochloric acid molarity which can gives the
highest retention of 68Ga in three different types of cartridges; determine the most suitable anion-exchange
column that can be implemented into the SnO2-based 68Ge/68Ga generator
MATERIALS AND METHODS
This research mainly involved laboratory works. Starting from synthesizing the Ga-68 until the last steps would
be the desorption step. Proper protection and safety while handling the radioactive compound would be taking
under full consideration and pre-caution. Particular attention would be paid at reducing the personnel radiation
exposure during the manual operation steps.
A 30 mCi, 68Ge/ 68Ga generator used was SnO2 Based Column generator manufactured by iThemba LABS
(Somerset West, South Africa) was used in this study. The column was loaded with the parent 68Ge (half-life
271 days). At this time, the generator capability of eluting Ga-68 is 16 mCi.
There are three different types of anion exchange columns were investigated in the study that are an in-jouse
made cartridge using AG 1-X8 Resin (250 mg, 200-400 dry mesh, 45-106 µm wet bead, BioRad, USA), Oasis®
WAX (30mg, Waters, USA) and Chromafix® PS-HCO3 column (45 mg, Machery-Nagel, Duren, Germany).
68Ga
at specific time. The time was taken for each reading in order to do the decay correct for the activity of the
Ga-68.
Deionized water was prepared by the Barnstead, Easypure II, RF ultrapure water system (USA). Basically, the
deionized was used for the dilution of HCl.
All the consumables that had been used such as tubing, connectors, syringes and needles are of one time-used
in order to prevent any cross contamination.
The generator was eluted with 10 ml of 0.6 M HCl. The first 1 ml fraction of 0.6 M HCl was eluted into the
waste vial followed by eluting next 2 ml of the 0.6M HCl into the collecting vial of 68Ga eluate. After that,
another 7 ml of the 0.6M HCl was eluted into the waste vial. For the safety purpose, the elution vial of 68Ga
was placed into lead pot.
Three different types of anion exchange columns were used for the adsorption of Ga-68 process. They were AG
1-X8 resin, Oasis WAX and Chromafix® PS-HCO3 columns. Before starting the adsorption process, all three
anion exchange columns were pre-conditioned with 2 ml 5M HCl and 10 ml of air to remove excess HCl within
the column. Then various molarities of Ga-68 eluate in HCl which ranged from 4.0 M to 7.0 will be investigated
to find the optimum HCl molarity which gave the higest 68Ga retention. Subsequently, 68Ga will be desorbed
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from the column with 1 mL of water for injection (WFI). Desorption was expressed as % 68Ga desorption of
total 68Ga applied to the columns. After each elution step, the anion column was flushed with air.
Various molarities of HCl which ranged from 4.0 M to 7.0 M HCl had been tested. The molarities of HCl were
4.0 M, 4.5 M, 5.0 M, 5.5 M, 6.0 M, 6.5 M and 7.0 M. Two mL of the 68Ga eluate was added to 4 mL 5.70 M
HCl and finally total volume would be 6 mL with final [HCl] was 4.0 M. In order to get 6 mL of 68Ga solution
with final [HCl] was 4.5 M, the 68Ga eluate was added to 4 mL 6.45 M HCl. Then, 68Ga eluate was added to 4
mL 7.20 M HCl and the final solution would be 6 mL with final [HCl] became 5.0 M. In order to produce 6 mL
68Ga solution with final [HCl] was 5.5 M, the two mL eluate which was eluted from generator was added to 4
mL 7.95 M HCl. After that, the next two mL eluate was added to 4 mL 8.70 M HCl in order to obtain 68Ga in
6 mL 6 M HCl. Then, two mL 68Ga eluate was added to 4 mL 9.45M HCl and final solution would be 6 mL
with 6.5 M HCl. Finally, two mL 68Ga eluate was added to 4 mL 10.20 M HCl in order to get the final 68Ga
solution would become 6 mL with 7.0 M HCl.
-
Gallium-68 activity in each different molarity as mentioned in the previous section was eluted through the
anion exchange columns. There were four readings were taken throughout the experiments. The parameters
were syringe activity of 68Ga before the elution, syringe activity of 68Ga after the elution, 68Ga activity in the
column after the elution and the 68Ga activity in the waste vials. Those parameters were measured by using
. This is due to the
decay process of the 68Ga activity. Therefore, by taking the time, the researcher was able to do the decay
correct in order to obtain an accurate activity of 68Ga.
The percentage of 68Ga retention formula as shown below: Equation (1)
Activity of 68Ga in anion column after elution x 100%
Syringe activity of 68Ga before the elution - Syringe activity of 68Ga after the elution
The researcher compared the percentage of 68Ga retention profiles between the three different anion cartridges
by taking the reading of %68Ga retention at the best molarity of HCl for retention process.
-
After the elution of 68Ga activity in various molarities of HCl as mentioned above, the adsorbed 68Ga in anion
exchange column was desorbed with one (1) mL of water for injection. The parameters that had been measured
are 68Ga activity in anion exchange column before desorption process, 68Ga activity in elution vial and 68Ga
activity in anion exchange column after desorption process. . The parameters were measured by using
The percentage of 68Ga desorption formula as follow: Equation (2)
68Ga activity in elution vial x 100%
68Ga activity in anion exchange column before desorption
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In this procedure, the researcher tried to identify whether the molarity of HCl may affect desorption of 68Ga or
not.
Statistical analysis was performed using IBM SPSS Statistics Version 20 software. In this study, data was
analyzed by using one-way ANOVA test. Statistical significance was defined at p<0.05.
RESULTS
All of data were collected from twenty one samples for each anion columns. In this research, three different
types of columns were investigated. Therefore, the total of samples was sixty three which consists of three
samples for each molarity of hydrochloric acid. The results and data that had been collected were percentage
of 68Ga adsorption and percentage of 68Ga desorption. Generally, the data obtained were different for diverse
HCl molarities. In this study, all the activity of gallium-68 was in millicurie unit (mCi).
-
68Ga activity in fraction of 6 mL at various [HCl] was measured by using dose calibrator and time was taken in
order to make decay correction of gallium-68 activity.
-
In the AG 1-X8 resins cartridge, a range of HCl molarities was run to determine which molarity that can give
the highest retention of gallium-68. The 68Ga adsorption in the column was expressed as percentage of 68Ga
retention. For identification of the best molarity which would give the highest 68Ga retention, test that had
been used was one-way ANOVA analysis due to more than two groups (seven groups) involved in the
experiments. There were seven groups in the experiment which is seven different molarities were tested.
Percentage of Ga-68 adsorption results for AG 1-X8 resin cartridges are shown in Figure 1. Based on the line
graph, the percentage of Ga-68 adsorption (retention) for AG 1-X8 columns begin to be plateau at HCl
molarity of 5.5.
Percentage of Ga-68 adsorption 100 83.74 86.66 90.67 97.07 96.72 96.69 96.77
90
80
70
60
50
40 Average %Ga-68
Adsorption
30
20
10
0
4.0 4.5 5.0 5.5 6.0 6.5 7.0
HCl Molarity
Figure 1: Graph of percentage of Ga-68 adsorption on AG 1-X8 column with different HCl
molarities.
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No. 2, December 2015.
By using one-way ANOVA, the analysis of data is shown in Table 1. The significance level for this study is
0.05 (p<0.05). It means that the mean difference is significant at the 0.05 level. Therefore, percentage of 68Ga
retention in AG 1-X8 column at 5.5M HCl is significantly different compared to 4.0 M, 4.5 M and 5.0 M.
However, the percentage of 68Ga retention at 5.5M, 6.0 M, 6.5 M and 7.0 M are equal. The equal variances
assumed by using Bonferroni.
-
In the Chromafix cartridge, a range of HCl molarities was run in order to determine which molarity that can
give the highest retention of gallium-68. The 68Ga adsorption in the column was expressed as percentage of 68Ga
retention. Data normality was tested by using Kolmogorov-Smirnov test. For identification of the best molarity
which would give the highest 68Ga retention, test that had been used was one-way ANOVA analysis due to
more than two groups (seven groups) involved in the experiments. There were seven groups in the experiment
which is seven different molarities were tested.
Percentage of Ga-68 adsorption results for Chromafix® PS-HCO3 columns are shown in Figure 2. Based on the
line graph, the % Ga-68 adsorption (retention) for Chromafix column begins to plateau at HCl molarity of 5.5.
Table 1: One-way ANOVA data analysis of percentage of Ga-68 retention in AG 1-X8 anion
columns at 5.5M compared to other HCl molarities
(I) (J) Mean Std. Sig. 95% Confidence
Molarit Molarit Difference (I- Error
y y Interval
J)
Lower Upper
Bound Bound
5.5M 4.0M 13.32667* 2.46562 .002 4.2058 22.4476
.018
4.5M 10.40667* 2.46562 .444 1.2858 19.5276
1.000
5.0M 6.40000 2.46562 1.000 -2.7209 15.5209
1.000
6.0M .34667 2.46562 -8.7742 9.4676
6.5M .38000 2.46562 -8.7409 9.5009
7.0M .29333 2.46562 -8.8276 9.4142
*. The mean difference is significant at the 0.05 level.
Percentage of Ga-68 adsorption 100 93.33 94.13 95.07 98.3 98.17 98.24 98.17 Average %Ga-68 adsorption
90
80
70
60
50
40
30
20
10
0
4.0 4.5 5.0 5.5 6.0 6.5 7.0
HCl Molarity
Figure 2: Graph of percentage of Ga-68 adsorption on Chromafix® PS-HCO3 columns with
different HCl molarities.
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No. 2, December 2015.
By using one-way ANOVA, the analysis of data is shown in Table 2. The significance level for this study is
0.05 (p<0.05). It means that the mean difference is significant at the 0.05 level. Therefore, percentage of 68Ga
retention in Chromafix® PS-HCO3 column at 5.5M HCl is significantly different compared to 4.0 M, 4.5 M and
5.0 M. However, the percentage of 68Ga retention at 5.5M, 6.0 M, 6.5 M and 7.0 M are identical. The equal
variances assumed by using Benferroni.
Table 2: One-way ANOVA data analysis of percentage of Ga-68 retention in Chromafix® PS-
HCO3 column at 5.5M compared to other HCl molarities
(I) (J) Mean Std. Sig. 95% Confidence
Molarit Molarit Difference (I- Error
y y Interval
J)
Lower Upper
Bound Bound
5.5M 4.0M 4.97333* .35764 .000 3.6503 6.2963
4.5M 4.16667* .35764 .000
5.0M 3.23333* .35764 .000 2.8437 5.4897
6.0M .35764 1.000
6.5M .13333 .35764 1.000 1.9103 4.5563
7.0M .06333 .35764 1.000
.13333 -1.1897 1.4563
-1.2597 1.3863
-1.1897 1.4563
*. The mean difference is significant at the 0.05 level
In the Oasis® WAX cartridge, a range of HCl molarities was investigated in order to determine which molarity
may give the highest retention of gallium-68. The 68Ga adsorption in the column was expressed as percentage of
68Ga retention. Data normality was tested by using Kolmogorov-Smirnov test. For identification of the best
molarity which would give the highest 68Ga retention, test that had been used was one-way ANOVA analysis
due to more than two groups (seven groups) involved in the experiments. There were seven groups in the
experiment which is seven different molarities were tested.
Percentage of Ga-68 adsorption results for Oasis® WAX columns are shown in Figure 3. Based on the line
graph, the % Ga-68 adsorption (retention) for Oasis® WAX column begins to plateau at HCl molarity of 5.5.
Percentage of Ga-68 adsorption 100 98.3 97.97 98.06 98.4
86.38
90
80 68.28 73.04
70
60
50
40 Average %Ga-68
30 adsorption
20
10
0
4.0 4.5 5.0 5.5 6.0 6.5 7.0
HCl Molarity
Figure 3: Graph of percentage of Ga-68 adsorption on Oasis® WAX columns with different
HCl molarities.
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By using one-way ANOVA, the analysis of data is shown in Table 3. The significance level for this study is
0.05 (p<0.05) which is the mean difference is significant at the 0.05 level. Therefore, percentage of 68Ga
retention in Oasis® WAX column at 5.5M HCl is significantly different compared to 4.0 M, 4.5 M and 5.0 M.
However, the percentage of 68Ga retention at 5.5M, 6.0 M, 6.5 M and 7.0 M are identical. The equal variances
assumed by using Bonferroni.
Table 3: One-way ANOVA data analysis of percentage of Ga-68 retention in Oasis® WAX
column at 5.5M compared to other HCl molarities
(I) (J) Mean Std. Sig. 95% Confidence
Molari Molari Difference Error
ty ty Interval
(I-J)
Lower Upper
Bound Bound
5.5M 4.0M 31.02333* 3.06627 .000 19.6805 42.3662
.000
4.5M 25.25667* 3.06627 .035 13.9138 36.5995
1.000
5.0M 11.91667* 3.06627 1.000 .5738 23.2595
1.000
6.0M .32667 3.06627 -11.0162 11.6695
6.5M .24333 3.06627 -11.0995 11.5862
7.0M -.09667 3.06627 -11.4395 11.2462
*. The mean difference is significant at the 0.05 level.
The descriptive analysis of % Ga-68 retention at 5.5 M HCl for those investigated anion cartridges are shown in
Table 4.
Table 4: Descriptive analysis of %Ga-68 retention at 5.5 M HCl
Types of Anion N Min Max Mean Std.
Cartridges Deviation
AG 1-X8 3 96.40 97.90 97.07 0.764
Chromafix 3 98.10 98.60 98.30 0.264
Oasis 3 97.50 99.10 98.30 0.800
Based on the results analysis, at 5.5 M HCl, the highest %Ga-68 retention is Chromafix and Oasis cartridge
which is 98.30% and the lowest retention is in AG 1-X8 which is only 97.07%.
-
For the desorption procedures, 1 mL of water for injection (WFI) was used in order to eluate the retained 68Ga
in anion columns into final vial. The highest average of %68Ga desorption is Oasis WAX which is 70.49%
followed by second highest is Chromafix which is 70.36% and the lowest is AG 1-X8 which is 58.56%.
Details of the results are shown in Table 5 and Figure 4.
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No. 2, December 2015.
Column Table 5: Average of %Ga-68 desorption in three different cartridges
AG 1-X8
Chromafix Average % 68Ga desorption (%)
Oasis WAX 58.56
70.36
70.49
80 70.36 70.49
70
60 58.56
Ga-68 Desorption (%) 50
40
30
20
10
0 Chromafix Oasis WAX
AG 1-X8
Figure 4: Average %Ga-68 desorption on AG 1-X8 Resin, Chromafix and Oasis WAX columns
DISCUSSION
68Ge/68Ga generators are frequently delivered with a given activity of 68Ge on the column, however, the
elutable amount of 68Ga activity is not stated and depends, among others on the carrier and eluate. For most
commercially available TiO2-based 68Ge/68Ga generators holds, e.g. 1000MBq 68GeCl4 on the column, and
7900MBq 68Ga can be eluted shortly after production, and after 1 half-life of 68Ge the elutable 68Ga activity is
reduced to 7350 MBq, thus 770% (Breeman et al., 2005; Decristoforoetal., 2007). iThemba Labs delivers their
SnO2-based generator with a given elutable 68Ga activity, which is >100% in the beginning and ±75% after
1half-life of 68Ge (E. de Blois et al., 2011).
-
Since the trapping of gallium-68 activity has potential to be increased and improved by using anion exchange
column, researcher attempted to compare the ability of three different cartridges which are AG 1-X8 Resin, 500
g, Chloride, 200-400 dry mesh, 45-106 µm wet bead (BioRad, USA) in an in-house made cartridge, Oasis®
WAX (30mg, Waters, USA) and Chromafix® PS-HCO3 column (45 mg), Machery-Nagel, Duren, Germany.
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No. 2, December 2015.
In order to compare the ability of anion exchange cartridge in term of purification, two aspects were focused in
this study which are percentage of gallium-68 retention in the columns and also the percentage of gallium-68
desorption on all the investigated cartridges. However, for identification of the percentage of Ga-68 retention,
different molarities of hydrochloric acid were used. According to Meyer et al. (2004), at [HCl] >5.0 M, Ga3+
ions will form the negatively charged gallium tetrachloro complex [GaCl4]- which can be used advantageously to
concentrate the eluted gallium , since the [GaCl4]- complex can be adsorbed quantitatively on a strong anion
exchange column. Therefore, in order to identify the best and most suitable molarities, researcher used a range
of different HCl molarities which are 4.0M, 4.5M, 5.0M, 5.5M, 6.0M, 6.5M and 6.0M of HCl.
In this study, researcher investigated the effect of molarity on the percentage of Ga-68 retention and finally
attempted to identify the most suitable HCl molarity for adsorption of Ga-68 purposes.
Based on the result analysis, in all of investigated cartridges, the percentage of 68Ga retention or adsorption
started to be plateau at molarity of 5.5M. Based on the One-way ANOVA analysis, there is no significant
difference between 5.5M with 6.0M, 6.5M and 7.0M which means that the retention of gallium-68 is equal at
those molarities. Therefore, 5.5M is the best molarity for the adsorption of gallium-68 in all three anion
exchange cartridges. This shows that at HCl molarity of 5.5M, all Ga3+ ions had form the negatively charged
gallium tetrachloro complex [GaCl4]- ; thus all of Ga3+ ions is quantitatively adsorb on the anion column while
other metallic impurities such as Germanium-68, Zinc-68, Stannous and other metallic impurities will be pass
through the column. Therefore there is no need to increase HCl molarity more than 5.5M in order to adsorb all
of the Gallium ion on the anion column because later the high H+ ion need to be neutralize by an equal amount
of buffer. So it would be sufficient to have an optimum amount of HCl molarity in order to trap the Gallium-68
ion so that it will become safer and easier for later manipulation before it can be applied for clinical used.
According to the descriptive analysis at 5.5M for all of the three cartridges, Chromafix and Oasis WAX
cartridges showed the highest retention of gallium-68 which is 98.30%. The lowest Ga-68 retention was gained
by AG 1-X8 column which is 97.07%. There seems to be no difference between both Chromafix and Oasis
WAX cartridges in the retention of Gallium-68.
In desorption process, one milliliter of water for injection was used in order to elute the retained Ga-68 activity
in the anion exchange resin. Regarding to the analysis of results, the highest percentage of 68Ga desorption was
obtained by using Oasis WAX cartridges which is 70.49% followed by Chromafix which is 70.36%. The lowest
desorption of gallium-68 was obtained by using AG 1-X8 column which is only 58.56%.
This is in agreement with the study done by Eric et al (2010), where the highest desorption was also in Oasis
WAX 30mg column. The desorption of Gallium-68 from the anion column is concentrated in one (1) ml which
yielded a high specific activity with less H+ ion than the fractionation method where the Gallium-68 ion is in 2
ml fraction of 0.6M HCl.
By considering all of aspects, the best anion exchange column that should be implemented for a SnO2 Based
68Ge/68Ga generator is Oasis® WAX (30mg, Waters, USA).
Although there are no significance difference in terms of adsorption of Ga-68 ion at 5.5 M HCl on both Oasis
WAX and Chromafix column which are 98.30%, the desorption profile of the Oasis WAX column is high
compared to other column which are 70.49%. Therefore, from this study, the most suitable cartridge and HCl
molarity that should be applied in purification and concentration of Gallium-68 eluate from a SnO2 Based
68Ge/68Ga generator is Oasis® WAX column with a HCl molarity of 5.5M respectively.
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No. 2, December 2015.
CONCLUSION
This study gave an important info on the methods to be used in the preparation of gallium-68 elution. By
identifying the best molarity and anion exchange column for the retention of Ga-68, the activity in the
collecting or final vial can be increased effectively. Other than that, it also can help to optimize the uses of each
Ga-68 from the 68Ge/ 68Ga generator. Several limitations occurred throughout this study. The first limitation is
all the purified and concentrated Gallium-68 from the anion cartridges ion shall be assessed their metal
contamination by conducting Atomic Absorption Spectrophotometry so that the reduction of metal
contaminants can be seen by adopting the anion purification method. Other than that, the study is limited to
the ability of 68Ge/68Ga generator to elute the Ga-68 for every five hours.
ACKNOWLEDGEMENTS
The authors wish to thank Malaysian Nuclear Agency and Nuclear Medicine Department of National Cancer
Institute Putrajaya, Ministry of Health for providing facilities for this research.
REFERENCES
B´e MM, Duchemin B, Browne E, Chechev V, Helmer RG, Schonfeld E. Table of Radionuclides: Comments on
Evaluations. ISBN 2 7272 0211 3. Paris, France: CEA/DIMIRI/LNHB; 1999.
Breeman, W.A.,DeJong,M.,Visser,T.J.,Erion,J.L.,Krenning,E.P.,2003.Optimising conditions
specific
forradiolabellingofDOTA-peptideswith 90Y, 111In and 177Lu athigh
activities.Eur.J.Nucl.Med.Mol.Imaging30,917 920.
Breeman, W.A.,deJong,M.,deBlois,E.,Bernard,B.F.,Konijnenberg,M.,Krenning, E.P.,
2005.RadiolabellingDOTA-peptideswith 68Ga. Eur.J.Nucl.Med.Mol. Imaging 32,478 485.
Decristoforo C, Knopp R, von Guggenberg E, Rupprich M, Dreger T, Hess A, et al. A fully automated
synthesis for the preparation of 68Galabelled peptides.Nucl Med Commun 2007;28:870 5.
Erik de Blois, Ho Sze Chan, CkiveNaidoo, et al. Characteristics of SnO2-Based 68Ge/68Ga Generator and
Aspects of Radiolabelling DOTA-Peptides. Applied Radiation and Isotopes 2010.
IDB Holland bv. 2011. Products / Gallium-68 Generator / Product details Gallium-68 generator. (online)
http://www.idb-holland.com/products/7/15_product_details_gallium-68_generator.html (10 February 2011).
K. P. Zhernosekov, D. V. Filosofov, R. P. Baum, P. Aschoff, H.-J. Adrian, H. Bihl, A.A. Razbash, M. Jahn, M.
Jennewein, F. Rösch, Processing of generator produced 68Ga for medical application.
Konstantin P. Zhernosekov, Dimitry V. Filosofov, Richard P. Baum, et al. Processing of Generator-Produced
68Ga for Medical Application. J Nucl Med 2007.
Lambrecht R, Sajjad M (1988) Accelerator derived radionuclide generators. Radiochimica Acta 43:171 179.
-68. J. Nucl. Med. 21, 171 173.
Mattia Asti, Giovanni De Pietri, Alessandro Fraternali, et al. Validation of 68Ge/68Ga generator processing by
chemical purification for routine clinical application of 68Ga-DOTATOC. Nuclear Medicine and Biology
2008;35:721-724.
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Marchol M. Ion Exchangers in Analytical chemistry: Their Properties and Use in Inorganic Chemistry. Prague,
Czech Republic: Academia; 1982.
Meyer, G.J.,Macke,H.,Schuhmacher,J.,Knapp,W.H.,Hofmann,M.,2004. 68Ga-labelled DOTA-
derivatisedpeptideligands.Eur.J.Nucl.Med.Mol.Imaging31, 1097 1104.
Mirzadeh S, Lambrecht R (1996) Radiochemistry of Germanium. J Radioanal Nucl Chem 202:7 102.
Modlin IM, Lye KD, Kidd M. A 5-decade analysis of 13,715 carcinoid tumors. Cancer. 2003;97;934-959.
N.S Loktionova, A. Belozub, et al. Purification of 68Ga Combining Cation and Anion Exchange Process.
P. Gebhardt, Th. Opfermann, H.P. Saluz. Computer controlled 68Ga milking and concentration system. Applied
Radiation and Isotopes 2010.
Rösch F, Knapp R (2003) Radionuclide generators. In: Rösch F (ed) Radiochemistry and Radiopharmaceutical
Chemistry in Life Sciences. Kluwer Academic Publishers, Dordrecht/Boston/London, p 81 118.
Strelow FWE, Victor AH, van Zyl CR, Eloff C. Distribution coefficients and cation exchange behaviour of
elements in hydrochloric acid-acetone. Anal Chem 1971;43:870 6.
TanveerHussainBokhari, A. Mushtaq, Islam Ullah Khan. Concentration of 68Ga Via Solvent Extraction.
Applied Radiation and Isotopes 2009;67,100-102.
Velikyan I, Beyer GJ, Langstrom B. Microwave-supported preparation of 68Ga bioconjugates with high specific
radioactivity. BioconjugChem 2004;15:554 60.
Zhernosekov KP, Filosofov DV, Baum RP, Aschoff P, Bihl H, Razbash A, et al. Processing of generator-
produced 68Ga for medical application. J Nucl Med 2007;10:1741 8.
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1 Physics with Electronic Department, Faculty of Science and Natural Resources, University of
Malaysia Sabah, 88450 Kota Kinabalu, Sabah, Malaysia.
2 School of Physics and Materials, Faculty of Applied Sciences Universiti Teknologi MARA, 40450
Shah Alam, Malaysia.
*Corresponding author: [email protected]
ABSTRACT
The most sensitive part of a metal oxide semiconductor (MOS) structure to ionizing radiation is the
oxide insulating layer. When ionizing radiation passes through the oxide, the energy deposited creates
electron/hole pairs. Oxide trapped charge causes a negative shift in capacitance-voltage (C-V)
characteristics. These changes are the results of, firstly, incre asing trapped positive charge in the
oxide, which causes a parallel shift of the curve to more negative voltages, and secondly, increasing
interface trap density, which causes the curve to stretch-out.
ABSTRAK
The most sensitive part of a metal oxide semiconductor (MOS) structure to ionizing radiation is the
oxide insulating layer. When ionizing radiation passes through the oxide, the energy deposited creates
electron/hole pairs. Oxide trapped charge causes a negative shift in capacitance-voltage (C-V)
characteristics. These changes are the results of, firstly, increasing trapped positive charge in the
oxide, which causes a parallel shift of the curve to more negative voltages, and secondly, increasing
interface trap density, which causes the curve to stretch-out.
Keywords: ionizing, MOS, radiation damage, silicon dioxide (SiO2)
INTRODUCTION
The radiation damage in the silicon dioxide (SiO2) layers consists of three components: the buildup of trapped
charge in the oxide, an increase in the number of interface traps, and an increase in the number of bulk oxide
traps (Oldham and McLean, 2003; Haider and Abdulah, 2009). Electrons and holes are created within the SiO2
by the ionizing radiation or may be injected into the SiO2 by internal photoemission from the contacts. These
carriers can recombine within the oxide or transport through the oxide. Electrons are very mobile in SiO2 and
move quickly to the contacts (Tamaki et al., 2009); in contrast the holes have a very low effective mobility
and transport via a complicated stochastic trap-hopping process (Li and Nathan, 2004). Some of these holes
may be trapped within the oxide, leading to a net positive charge. Others may move to the SiO2/Si interface,
where they capture electrons and create an interface trap. Along with the electron-hole generation process,
chemical bonds in the SiO2 structure may be broken (Street, 2005). Some of these bonds may reform when the
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electrons and holes recombine, whereas others may remain broken and give rise to electrically active defects.
These defects can then serve as interface traps, or trap sites for carriers.
Bonds associated with hydrogen or hydroxyl groups can release these impurities when broken, which are then
mobile within the SiO2. These impurities may then migrate to the SiO2/Si interface, where they undergo a
reaction which results in an interface trap. The defects created by the radiation may themselves migrate in the
strained region near the SiO2/Si interface and also result in the formation of an interface trap. Typically the
net charge trapped in the oxide layer after irradiation is positive. The interface traps can exchange charge
freely with the silicon substrate, and thus their charge state depends upon the bias applied to the device, it is
more negative for a positive bias applied to the gate electrode than a negative bias. The energy band diagram
of an MOS structure under positive gate bias is as shown in Fig. 1. There are three possible oxide trap
locations.
Figure 1. Energy band diagram of a MOS structure under positive gate bias
RADIATION TEST PROCEDURE
To study the radiation damage in the SiO2 layer, 14 MeV neutrons produced from the deuterium-tritium (D-T)
nuclear fusion reaction was utilized. Fusion reactions are highly exothermic reactions, usually between light
nuclei, such as the isotopes of hydrogen, viz., deuterium (D=2H) and tritium (T=3H). The high voltage power
supplies used Felici type high-voltage multipliers for accelerating light nuclei induced reaction (IAEA, 1996;
Pereda et al., 2008). The low voltage (100 kV) neutron generators produced neutrons through the following
reaction, 2H 3H 24He n (Q = 17.6 MeV) or can be written as 3H (d, n)4He .
The 4He and n share 17.6 MeV consistent with linear momentum conservation, and a monoenergetic neutron
with energy 14 MeV emerges. This reaction often serves as a source of fast neutrons. The parameters of
SAMES J-25 as a 14 MeV neutron source are listed in Table 1.
Table 1. Parameters of Generator Neutron Type SAMES J-25.
Beam energy 150 KeV (optimal)
D+ Beam current at fix target 1.50 mA
Min. beam spot size at the target 3.10 cm
D-T neutron yield, continuously 3 109 n/s (optimal)
Target life (half value of neutron yield) 200 mAh. (expected)
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More than 25 GF4936 dual n-channel depletion mode MOS samples were irradiated at room temperature with
D-T neutrons to examine the neutron induced changes in the SiO2 of the MOS devices. All samples were set
and tested around the target at different angles, that is, (0o, 15o, 30o, 60o, and 90o). The differences in angles
were related to the total fluence of the neutron radiation received and its energy. Fig.2 shows the location of
sample during neutron irradiation.
Figure 2. Location of sample toward neutron radiation.
Each irradiation of samples was performed twice; first to measure the output characteristics, and second to
measure the forward transfer characteristics. In output characteristics, the gate-source voltage, VGS, was
changed from +0.5 to -1.5 volts by changing the applied DC supply voltage, Vcc, starting from 0 to 20 volts.
On the other hand, the characterization of MOS in order to determine the forward transfer characteristics of
the gate-source voltage, VGS, was changed from 5.0 to +5.0 volts with Vcc constant of 20 volts during
irradiation time.
RESULTS AND DISCUSSION
While the holes generated under exposure to ionizing radiation, travel through the oxide, MOS structures
typically exhibit a negative voltage shift, Vot, that is not sensitive to silicon surface potential and that can
persist for hours to years. This long-lived radiation effect component is the most commonly observed form of
radiation damage in MOS devices and is attributed to the long-term trapping of net positive charge in the
oxide layer as shown in Fig. 3. This effect generally dominates other radiation damage processes in MOS
structures, including negative charge (electron) trapping and interface trap buildup effect, unless specific
device-processing changes are made to alter the oxide and consequently reduce the positive charge trapping or
enhance the other effects.
Figure 3. Oxide charge (hole) trapping and removal processes in a MOS structure under
positive gate bias.
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Most of the holes that are captured by hole traps within a certain distance of the silicon (2-5 nm) are believed
to undergo a tunneling process (tunnel anneal) and are removed gradually, this process is thought to be
primarily responsible for the "long-term annealing" of Vot. Meanwhile, electrons that are generated in the oxide
within the trapped-hole distribution, or between the trapped-hole distribution and the silicon, swept through
trapped-hole distribution. Depending upon the local density of trapped holes, nht and the cross-section for
capture of an electron by a trapped hole, r some of the electrons recombine with a portion of the trapped
holes. Another possible process is of free electrons with free holes in the bulk of the oxide. This process is
distinct from germinate or columnar recombination of electrons and holes during the charge generation process,
and is not significant at reasonably attainable charge generation (dose) rates. The hole generation and trapping
processes produce the buildup of Vot under irradiation; the electron/trapped-hole recombination and hole
anneal processes tend to limit or reduce Vot. The overall charge buildup process (neglecting hole-removal
processes other than recombination) can be expressed in the following incremental form (Arora, 2007):
nht (x, D) = Fh(x)ht ( ox (x()[Nht (x) - nht (x)] - Fe (x) r ox (x))nht(x) . (1)
where
nht is the local density of trapped holes
N ht is the local density of hole traps
r is the local oxide field-dependent cross-section for recombination of an electron with a trapped hole
Fh and Fe are the local fluences per unit dose for radiation generated holes and electrons, depend upon
the "upstream" generation and removal of holes and electrons in the oxide
insert space
TheVot Vot that results from the trapped holes is given by
Vot = -(q/ ox ) doxNot . (2)
where Not Notis the areal charge density referred to the SiO2/Si interface and is given in turn by
N ot= 1 d o x nht (x)xdx . (3)
dox 0
Under the assumption that holes trapping takes place very close to the SiO2/Si interface in an MOS structure
irradiated under positive bias. The efficiency of hole captured by the traps is a function of the electric field in
the oxide. Another important parameter is the cross-section, r, for capture of a free electron by a hole in the
hole trap. The process of electron/hole recombination via the hole traps plays a major role in the limitation of
hole trapping at higher doses (Oldham and McLean, 2003).
The holes created and trapped in deep traps in SiO2 layer of a MOS structure after irradiations are not truly
"permanently" trapped. Instead, they are observed to disappear from the oxide over times from milliseconds to
years. This discharge of the hole traps, as commonly observed at or near room temperature, is the major
contributor to the so-called long-term annealing of radiation damage in MOS devices. The annealing of the
trapped holes has two manifestations that may reflect different hole-removal processes. The first is the slow
bias-dependent recovery of Vot typically observed at normal device operating temperatures (-55 to 125C, for
instance). The second is the relatively rapid and strongly temperature-dependent removal or recombination of
the holes observed when MOS structures are deliberately subjected to thermal annealing cycles at elevated
temperatures (150 to 350C).
The electrons generated by irradiation are much more than holes, and they are swept out of the oxide in times
typically in the region of 1 ps. However, in that first picosecond, some fraction of the electrons and holes will
recombine. This fraction depends greatly on the applied field and the energy and the type of incident particle.
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The holes that escape initial recombination are relatively immobile and remain behind near their point of
generation, causing negative voltage shifts in the electrical characteristics of MOS devices. However, over a
period of extended time at room temperature, the holes undergo a rather anomalous stochastic hopping
very dispersive in time, gives rise to a short term, transient recovery in the voltage shift. It is sensitive to many
variables including primarily applied field, temperature, and oxide thickness (Oldham, 2011). When the holes
reach the SiO2 interface (for positive applied gate bias), some of them are captured in long term trapping sites,
and cause a remnant negative voltage shift.
Ionizing radiation induced positive charges (holes) in the insulator SiO2, will then require a greater negative
turn-on voltage, or threshold voltage. Fig. 4 shows the kinds of changes that occurred in threshold voltage for
n-channel and p-
Figure 4. The changes of the threshold voltages of n- and p-channel MOS
transistors as a function of fluence.
For the purpose of calculating the linear energy transfer (LET) in the device when = 0o, the following
relationship is used; LET = E (MeV) s (Chee et al., 2009), where is the silicon density ( 2.33103 mg/cm3)
and s is the tack length of energy deposited by the incident particle (1 pC/m), thus, the LET = 96.57
MeVcm2/mg. In order to calculate the LET at different angles, the following relationship is used; LET =
E(MeV) cos /c , where c is the device thickness (1 micron), and the results are 92.7, 83.04, and 57.93
MeVcm2/mg at 15o, 30o, and 60o angles of projection respectively. The vertical penetration induced the smallest
energy deposition. In contrast, the horizontal penetration gave the largest energy deposition. It has been
estimated that the level of displaced atoms expelled from lattice position, the microscopic total cross section for
14-MeV neutrons in silicon is D 4.8 barn . The corresponding neutron mean free path, is given in turn by
(N D )1 (51022 4.81024 )1 4cm . (4)
This can be interpreted as the probability that an individual incident neutron produces one primary per cm is
equal to 1/4. The number of primaries then produced per cm3 is n / 4 . Each primary yields about 500 total
displacements, so that ns 500 . Then, for a different fluence, the mean number of displaced atoms per cm3, Nd
is shown on Table 2, Table 3, Table 4, Table 5 and Table 6 respectively at the incidence angle of 0o, 15o, 30o,
60o, and 90o respectively.
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Table 2. The mean number of displaced atoms per cm3 at the incidence angle of 0o.
Ave. Fluence (n/cm2) Nd (atoms/cm3)
0.66 1010 0.82 1012
1.32 1010 1.65 1012
1.98 1010 2.47 1012
2.64 1010 3.30 1012
3.31 1010 4.10 1012
1.98 1010 2.46 1012
Table 3. The mean number of displaced atoms per cm3 at the incidence angle of 15o.
Ave. Fluence (n/cm2) Nd (atoms/cm3)
0.39 1010 0.48 1012
0.78 1010 0.97 1012
1.16 1010 1.45 1012
1.55 1010 1.93 1012
1.90 1010 2.30 1012
1.15 1010 1.42 1012
Table 4. The mean number of displaced atoms per cm3 at the incidence angle of 30o.
Ave. Fluence (n/cm2) Nd (atoms/cm3)
0.75 109 0.94 1011
1.51 109 1.88 1011
2.26 109 2.82 1011
3.00 109 3.77 1011
9.78 109 4.72 1011
2.26 109 2.80 1011
Table 5. The mean number of displaced atoms per cm3 at the incidence angle of 60o.
Ave. Fluence (n/cm2) Nd (atoms/cm3)
0.32 109 0.40 1011
0.64 109 0.81 1011
0.97 109 1.21 1011
1.29 109 1.61 1011
1.62 109 2.00 1011
0.96 109 1.20 1011
Table 6. The mean number of displaced atoms per cm3 at the incidence angle of 90o.
Ave. Fluence (n/cm2) Nd (atoms/cm3)
1.68 108 2.10 1010
3.36 108 4.20 1010
5.00 108 6.30 1010
6.72 108 8.40 1010
8.40 108 10.00 1010
5.03 108 6.20 1010
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CONCLUSION
When ionizing radiation passes through the oxide, the energy deposited creates electron-hole pairs. High-energy
neutron radiation deposits energy in MOSFET device via two mechanisms: trapped charge buildup in the
silicon dioxide layer and also an increase in the density of trapping states at the silicon dioxide interface. The
electric characteristics of the MOS structure are easily changed by charges trapped in the SiO2 passivation
layer. This means that the MOS devices are influenced more by ionization effects than displacement effects.
ACKNOWLEDGEMENT
The authors are thankful to Fundamental Research Grant Scheme (FRGS) 2013, Project No.: FRG0318-SG-
fusion of Charge-Carrier in Semiconductor and nanostructure Devices and
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Oldham, T. R., McLean, F. B; (2003); Total ionizing dose effects in MOS oxides and devices; IEEE
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Chee, F. P., Haider F. Abdul Amir, and Saafie Salleh; (2009); System design for enhanced in situ monitoring
of ionizing radiation effects in semiconductor devices; Borneo Science, Vol. 25, 2009.
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1Fakulti Sains dan Bioteknologi, Universiti Selangor, 45600 Bestari Jaya, Selangor Darul Ehsan,
next line
Malaysia.
2Centre for Nuclear Energy, College of Engineering, Universiti Tenaga Nasional, 43000 Kajang,
Selangor, Malaysia.
3Pusat Pengajian Sains Fizik, Universiti Sains Malaysia, 11800 Penang, Malaysia.
4Agensi Nuklear Malaysia, 43000 Kajang, Selangor, Malaysia.
ABSTRACT
Thermal neutron beam from thermal column was selected for a Boron Neutron Capture Therapy
(BNCT) system utilizing the Malaysian TRIGA MARK II reactor. Determination of shielding
materials for fast and epithermal neutron was conducted. The materials selected were polyethylene,
paraffin and water. For gamma-ray shielding, lead was used. The objective of this paper is to present
the simulation and verification of an optimal design of BNCT collimation at a beam line viewing the
thermal column. A collimator was made from polyethylene pipe with 8 cm of diameter filled with
paraffin.
ABSTRAK
Thermal neutron beam from thermal column was selected for a Boron Neutron Capture Therapy
(BNCT) system utilizing the Malaysian TRIGA MARK II reactor. Determination of shielding
materials for fast and epithermal neutron was conducted. The materials selected were polyethylene,
paraffin and water. For gamma-ray shielding, lead was used. The objective of this paper is to present
the simulation and verification of an optimal design of BNCT collimation at a beam line viewing the
replace abstrak:
thAerlumralnecoulutrmonn. tAermcolalimdaatroirtuwraussmteadrme farotmelapholydeitphiylliehneunptiupek wpiethm8bacnmguofnadinamseistetremfilleBdowroitnh
paNraefufitnr.on Capture Therapy (BNCT) di reaktor penyeldikan TRIGA MARK II di Malaysia.
PKeenyewnoerdnstu: aconmbpaoshitae,nthpeermmoegrriaspihaya,nraudn-wtuaskten, eimuptraoctnsltarejungdthan neutron epitermal telah
dijalankan. Bahan-bahan yang digunakan ialah polietelina, parafin dan air. Bagi
pemerisian sinar gama, plumbum telah digunakan. Objektif kertas kerja ini ialah untuk
mengemukakan simulasi dan penentusahan rekabentuk optima bagi pengkolimat
align paragrappBhaNipCTpopliaedtealeanraahdaelnugranneudtiraomnedtIeiNdreT8pRacOnmDtudUrauCnsTdtIeiiOsrmiNdae.nSgaatnu pengkolimat telah dibina dari
parafin.
A very promising cancer treatment called Boron Neutron Capture Therapy (BNCT) is selected to be
studied in this research. BNCT is a radiation therapy for the treatment of cancers like melanomadamanindistrating
glioblastoma multiforme (Yanagie et al., 2010). BNCT is done by firstly, a stable isotope of boron -10
(10B) is administered to the patient via a carrier drug and then the patient is irradiated with a neutron
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No.72,sDheoceumldberb2e015.
superscipt
beam. The 10B will then undergo the capture reaction 10B(n, α)7Li where 10B capture cross section for
thermal neutrons is 3840 barn (Valda et al., 2005). This is the scientific fact on why thermal neutron is
used in BNCT.
carried out
MATERIALS AND METHOD
Neutron flux measurements were done with Fission Chamber detector. Back wall of the shielding box had
a hole on it with 10.0 cm of diameter. A small polyethylene collimator was placed inside the shielding box to
reduce the beam size for safety precautions. Fission Chamber detector was then placed at the back of the
shielding box. Water tank was placed at the back of the Fission Chamber detector as biological shielding.
Neutron flux measurement was done at the reactor power of 100 kW for 1 hour.
align paragraph measured
Figure 1. Neutron Fission Chamber detector Figure 2. Neutron spectrometry detector
Thermal column Beam stopper
beam port
Detector poisition
Figure 3. Detection Set-up
RESULTS AND DISCUSSION
Neutron beam that came out from the hole of thermal column door was measured with Fission Chamber detector.
Table 1 shows the counts of neutron particles measured with the Fission Chamber detector. The measurement was
done in a period of one hour and at reactor power of 100 kW. The average of neutron count per second was 2.30 x 104
s-1. Neutron flux calculated from the neutron count for exposed area of 1.8 cm2 on the Fission Chamber detector was
1.35 x 104 cm-2s-1. Calculated neutron flux for 1 MW of reactor power was 1.35 x 105 cm-2s-1.
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No. 2, December 2015.
Table 1. Counts of neutron particles measured with the Fission Chamber detector.
Time in Seconds 600 1200 1800 2400 3000 3600 Average
Neutron per Seconds 2.30 2.40 2.40 2.40 2.40 2.20 2.30
(× 104)
Figure 4. Combination of all neutron spectrums thermal column beam port with and without
filter/moderator materials
CONCLUSION
The fluxes measured were very much lower than all measurements done with TLD detectors. This was believed
because of the aperture used in this measurement that only allowed a very small neutron beam (1.5 cm of beam
diameter). The measurement with TLDs use neutron beam with 8 cm of diameter. A bigger aperture will give
higher flux and will be sufficient to be used for BNCT treatment.
used
ACKNOWLEDGEMENT
Special thanks to Composite Technology Malaysia for providing the test materials. Our gratitude go also to
Universiti Tenaga Nasional, Universiti Sains Malaysia, Universiti Selangor, Hokkaido University and Nuclear
Malaysia for supporting this work.
REFERENCES
1. Yanagie H., Kumada H., Nakamura T., Higashi S., Ikushima I., Morishita Y., Shinohara A., Mitsuteru
F., Suzuki M., Sakurai Y., Sugiyama H., Kajiyama T., Nishimura R., Ono K., Jun N., Minoru O.,
Eriguchi M., and Takahashi H. (2010). Feasibility evaluation of neutron capture therapy for
hepatocellular carcinoma using selective enhancement of boron accumulation in tumour with intra-
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JOURNAL Of NUCLEAR And Related TECHNOLOGIES, Volume 12, No. 2, December 2015.
arterial administration of boron-entrapped water-in-oil-in-water emulsion. Appl Radiat Isot. Volume
69(12), pp.1854-1857.
2. Valda A., Minsky D. M., Kreiner A. J., Burlon A. A. and Somacal H. (2005). Development of a
tomographic system for online dose measurements in BNCT (Boron Neutron Capture Therapy).
Brazilian Journal of Physics, Volume 35, no. 3B. Pp.785-788.
93
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