FNCA Newsletter No.10
Radiation Safety & Radioactive Waste Management
C o nte nts
The 2015 RS&RWM Workshop
P.1 – P.2

March 2016

Topics from participating countries
P.3 – P.18

Issued by
Nuclear Safe ty
Research Association(FNCA Secretariat,
JAPAN )

List of Project Leaders P.19
FNCA Framework P.20

The 2015 FNCA Workshop on
Radiation Safety & Radioactive Waste Management (RS&RWM)
17‐19 November, 2015, Serpong, Indonesia

The FNCA 2015 Workshop on
Radiation Safety and Radioactive Waste
Management (RS&RWM) was heldfrom November 17 to 19 2015, in
Serpong, Indonesia. This workshop was
hosted by the Ministry of Education,
Culture, Sports, Science and Technology(MEXT) of Japan and the National
Nuclear Energy Agency of Indonesia(BATAN). Seventeen researchers andexperts from Bangladesh, China,
Indonesia, Japan, Kazakhstan, Malaysia,
Mongolia, Thailand and Vietnam
participated in the workshop.
On the first day, member countries
gave
acountryreporton
nuclear/radiologicalemergency
preparedness and response detailingpolicy, regulatory framework andlegislation, emergency classes andcondition, participating organization,
concept of operation such as mitigationand prevention, emergency preparedness,
emergency response and emergency

rehabilitation and restoration, and otherrelated activities such as human capacity
building and communication and public
relation.
Participants then reported progress of
theirconsolidated
reporton
nuclear/radiologicalemergency
preparedness and response and agreed tocomplete their first compiled draft by theend of March 2016.

Figure1: The 2015 RS&RWM Workshop

1system, compaction system, incineration
system, and cementation system.
On the last day of the workshop,
participants had a discussion on status,
plans and challenges of low/Intermediatelevel waste disposal facilities/long term
storage facilities. At the wrap up session,
participants exchanged their opinions
on subjects that they are interested
in the field of RS&RWM andactivities to be conducted in thenear future.

An open seminar on countermeasuresfor nuclear/radiological emergency at
nuclear facilities was held at BATAN
Serpong on 18 November. The opening
remarks were made by Prof Toshiso
Kosako, Emeritus Professor, The
University of Tokyo and Mr Suryantoro,
Head, Center for Radioactive Waste
Technology (CRWT), BATAN. Eight
presentations were provided on each
topic
of
countermeasuresfornuclear/radiologicalemergency
atnuclear facilities in Bangladesh,Indonesia, Kazakhstan, Malaysia, andThailand. Japanese participants gave apresentation on overview and recent
progress of FNCA, countermeasures
under Fukushima Daiichi NPP accident,
and human resource development
programs
of
Japan
for
nuclear/radiological emergency. A panel
discussion was held concerning the
topics of 1) sharing information onlessons
learnedfromnuclear/radiological accident, 2)
countermeasuresforexternal
events(natural disaster etc.), and3)human resources development.

Figure3: Center for Radioactive Waste Technology(CRWT), BATAN

Figure2: Open Seminar

A technical visit to the BATAN
Serpong was conducted on the afternoonof 18 November 2015. Participants
visited the GA. Siwabessy Research
Reactor, which is a 30 MW multipurpose
reactor and reached the first criticality in
1987 and the Center for Radioactive
Waste Technology (CRWT), which wasestablished as a centralized radioactive
waste management facility of Indonesia
in 1989. They observed evaporation

Figure4: GA. Siwabessy Research Reactor, BATAN

2

Topics from Participating Countries
There is no waste return to Australia
for spent fuel elements shipped to the
USA under the US National Nuclear
Safety Administration’s Global Threat
Reduction Initiative (GTRI). However
the inter-governmental agreements madewith France and the UK stipulates thereturn of wastes arising from thereprocessing of ANSTO’s spent fuelback to Australia. It is internationallyaccepted that countries be responsible
for the management of the radioactive
waste they produce.

Australia
Ms Lynn Tan
Australian Nuclear Science &Technology Organisation (ANSTO)

Return of Intermediate Level
Waste from France
The Australian HIFAR Reactor
The High Flux Australian Reactor(HIFAR) at Lucas Heights which went
critical on the 26 January 1958 is thefirst nuclear reactor in the southernhemisphere. It was used for producing
neutrons for scientific research andradioisotopes for medicine and industry.
HIFAR retired in January 2007 after
close to 50 years of reliable operation.

Reprocessing of Spent Fuel
Reprocessing is the chemical process
whereby fissile and fertile materials inthe spent fuel are recovered in order toprovide fresh fuel for existing and futurenuclear power plants. It offers benefits inincreased uranium and plutonium
utilisation, and a reduction in the volumeof high level waste.
The resulting liquid waste from thereprocessing process is then melted
together with glass material at up to
1300oC for incorporation into the glass
matrix. The melted mixture is poured
into stainless steel canisters which arewelded shut after controlled cooling. The
canisters undergo a wet decontaminationand pin shotting to remove possible
surface contamination prior to interim
storage or shipment.

HIFAR's reactor core consisted of 25
fuel elements and has a nominal
maximum thermal power output of 10
megawatts. Approximately 37 spent fuelelements (SFE) were generated by
HIFAR each year. As Australia does nothave a nuclear spent fuel reprocessing
facility, these SFE, following a period of
wet then dry storage, were sent overseas
for reprocessing.
Shipment of Spent Fuel
Between 1996 and 2009, there wereeight overseas shipments of used HIFAR
spent fuel - four shipments to France forreprocessing, three shipments to theUnited States (US) as part of the US
Department of Energy (DOE) Foreign
Research Reactor Spent Nuclear Fuel
Acceptance Program (FRR-SNF) andone shipment to the United Kingdom(UK) also for reprocessing. Eachshipment was carried out safely using
approved
andlicensedtransport
packages in dedicated INF-2 classified
ships.

Wastes Arising from Reprocessing inFranceA total of 1288 spent fuel elements(SFE) were sent to France in four
separate shipments between 1999 and2004. An equivalent amount of wastes
from the reprocessing of these 1288 SFE
was agreed upon for return to Australia.
(i) Vitrified Residues
A total of 20 steel canister of vitrified
intermediate level waste with less than
300 W/m3 of total thermal heat were3

assigned to Australia. These 20 canisters
were loaded into a French-designed anddeveloped dual purpose transport andstorage cask, also known as TN81. The
TN81 is classified as a Type-B transport
container under the international
regulations for the safe transport ofradioactive materials. It is designed towithstand a drop of 9 metres,
temperatures above 800oC, earthquakes
and jet plane crashes. In transportconfiguration the cask measures around7m in length and 2.7m in diameter, andwhen fully loaded weighs close to 120
tonnes. Containment and shielding isprovided by the thick forged steel
cylindrical vessel with a welded bottom,
the external lead-filled aluminium
profiles and the primary lid with theassociated seals.

waste packages were unloaded from theship and secured onto waiting trucks fordelivery to ANSTO’s Lucas Heights site,
approximately 60km north of the port.
All maritime and road transport
approvals were sought from and grantedby the relevant local, national andfederal regulatory bodies in France andAustralia. A comprehensive security
operation was also in place to ensure thatsafety and security were maintained
throughout the operation.

(ii) Technological Wastes
The technological wastes incidental tothe operation and maintenance of the
reprocessing
facility(suchascontaminated gloves and tools) were also
assigned pro rata to Australia. These arealso intermediate level wastes, thoughmuch less radioactive than the vitrified
residues.

Figure2: TN-81 Cask being loaded onto a Class INF2 ship

Operational Readiness Program
A comprehensive training program
was set up to prepare the project teamand the relevant stakeholders for thereceipt of the first waste return ofAustralia’s reprocessed spent fuel. The
training format included both classroom
style lectures and on-site practice withmock-up equipment. The training was
conducted in phases during the course of
the project, utilising subject matter
experts and teaching materials soughtfrom within Australia and overseas, anddrawing upon the experience of overseas
owners of similar casks.

Figure1: The cemented technological wastes are
cemented within overpacks and placed inside an ISO
IP2 transport container, DV78.

Shipment back to Australia
Both waste packages were loaded onto
an INF2 rated ship purpose-built for
nuclear material shipment, at Cherbourg,
France on October 2015 after a thoroughinspection of the vessel by the Frenchmaritime authorities.
The position of the ship was tracked
throughout its two months journey,
arriving in Australian waters on
December 2015. At the port, the two

Figure3: Hands-on training (a) Use of building crane tolift and tilt a mock-up TN81 cask from the transport
frame (b) Test run of leak tightness equipment set up ona mock up secondary lid.

4

The training covered a broad area,
including work health safety, radiation
safety, regulatory, engineering, waste
management, transport, security, safety
and reliability. It involved staff acrossmultiple disciplines and divisions within
ANSTO, as well as external commercial
partners, regulatory bodies and agencies.
The training provided a forum forinteraction and discussion amongst thevarious groups, and equipped theparticipants with knowledge and skills to
confidently plan and execute theirrespective roles and responsibilities
required by the project.

DV78
TN81

Figure4: (a) TN81 arriving into the IWS on the back of
the transport trailer. (b) DV 78 and TN81 inside the IWS.

A thorough radiation dose andcontamination survey was conducted on
both waste containers by the radiation
protection advisors and health physics
surveyors. No contamination was found.
The dose rates on the TN81 and DV78
packages measure 0.011mSv/h and
0.350mSv/h respectively, well below the
maximum permitted transport container
surface dose rate of 2 mSv/h according
to the IAEA transport regulations.

Arrival of Waste Packages at ANSTO
Due to the mass and size of the load,
road transport of the packages from theport to ANSTO was constrained to theearly morning hours and on the night ofleast traffic density to minimise impacts
on the public. The convoy arrived at theLucas Heights site in the early hours of
Sunday morning on 6th December 2015.

Setting up of TN81 for Interim
StorageIn order to store the TN81 in thefacility, a series of operations were
undertaken to convert the cask from its
transport configuration into a storage
configuration over the following few
days.
Day
Activities
1 Remove impact limiters. Tilt andlift from trailer. Locate withinscaffold.
2 Prepare secondary lid. Mount leaktightness monitoring equipment.
Evacuate inter‐lids space.
3 Conduct leak tightness test. Installanti‐aircraft crash cover.
4 Connect to SCADA system forcontinuous
monitoring
andrecording. Remove scaffold.

Figure3: TN81 being transported on ANSTO’s Lucas
Heights site.

A purpose-built an interim waste
storage facility, named the Interim Waste
Store (IWS), was licensed by theAustralian Radiation Protection and
Nuclear Safety Agency (ARPANSA) in2015 to receive and temporarily store thereturned waste from France. The
building is equipped with a 140T DG
rated crane.
The DV78 with the technologicalwaste was delivered first to the IWS andplaced into position using the building
crane. The TN81 was driven into thebuilding next on the back of a 7-axle
trailer.

Table 1: Program to set the TN81 into storage
configuration.

This included firstly the removal of thefront and back impact limiters. A twolegged lifting beam was then attached to
the front trunnions of the cask. Using the5was achieved and the minimum leak rate
was not exceeded.
On the fourth day, the TN81 wasconnected to the monitoring and logging
system within the IWS for continuous
monitoring and recording of the pressure
signals from within the inter-lids spaceof the TN81 and the temperature of the
IWS. The SCADA system is linked tothe site control centre for notification of
excursions from any alarm set points.

140T DGR building crane, the cask was
carefully tilted and lifted off thetransport frame on the back of the trailer.
The cask is then located in its storage
position and a scaffold built around it toallow access to the top of the cask.

Figure5: (a) Tilting and lifting of the TN81 from thetransport frame. (b) Handling team posing in front of thevertically positioned TN81.

On the second day, the secondary lid
was cleaned and fitted with two seals – ametallic seal and an elastomer seal. The
monitoring system consisting of apressure vessel and three pressuretransducers was also secured onto thesecondary lid before bolting the preparedlid onto the cask over the primary lid.

Figure7: TN81 in storage configuration inside the IWS.

This project completes a significant
milestone in the management of
ANSTO’s backend fuel cycle. It has
involved not only multi-disciplinary
participation and dedication across
ANSTO, but also close collaboration andcooperation with commercial partners,
industry experts, regulatory bodies andgovernmental bodies both nationally andinternationally.
National Waste Repository
The Australian National Radioactive
Waste Management Facility (NRWMF)
is anticipated to be operational by theend of the decade, at which time theFrench returned waste will betransported from the IWS in ANSTO for
long-term storage in the NRWMF.

Figure6: (a) Lowering of the secondary lid over the
primary lid of the TN81. (b) Positioning the bolts forsecuring the secondary lid.

After securing the secondary lid, theleak tightness equipment was set up andthe inter-lid spaces evacuated overnight.
On the third day, the inter-lid space
was systematically pressurised to 6.5 bar
absolute, first with nitrogen, then withhelium. This was followed by a leaktightness test to ensure the required seal

6

storage facility. The activated graphite
sample collected from research reactor
facility was analysed for strong gammalines by using HPGe detector coupledwith Genie software. A series of activity
measurements with a few representative
waste containers were performed. Before
the experiment took place a written
system of work was produced.

Bangladesh
Dr M. Moinul Islam
Bangladesh Atomic Energy
Commission (BAEC)

Gamma Analysis of Solid
Radioactive Waste Produced in
the Nuclear Facility

Experimental Methods
To aid risk assessment a non-active run
was initially performed, with a sample
for an unirradiated graphite sample.
Contamination was checked before
scraping. To avoid finger cuts fromsharp blades, latex gloves and leather
gloves were worn. In order to analyse thegraphite plug three representative
samples were taken from activatedgraphite. By using / survey meter the
maximum average dose rate and surface
activity of the sample was recorded 2.80
Sv/h and 8.48 Bq/cm2 respectively. A
scraping technique was performed totake/collect the sample from the active
graphite. Polythene, latex and leather
gloves were worn to avoid cuts andcontamination safety spectacles wereused to protect eyes from dust during the
scraping of the graphite samples.
Protective clothing, lab coats and masks
and pocket dosimeter, TLD were worn
before entering the waste storage facility
in order to collect the graphite sample.
Any loose particles from the graphite
were picked up on plastic tape. The
plastic tape was put into a 250 ml plastic
pot. The sample was then collected in theplastic containers. The sample was thenput into a lead pot to take up to thecounting room.

Introduction:
The decommissioning and dismantlingof nuclear facility after their service life
involves the necessary disassembling,
handling and disposing of a large amountof radioactive equipment and structures.
To reduce this undesirable volume to theminimum and to successfully plan thedismantling and disposal of radioactive
materials to storage facilities, theactivations of the structures should be
accurately evaluated. To optimize themass flows going to waste disposal it isimportant to determine the radionuclide
present in the waste materials generated
due to the long-term induced activity. In
particular, the graphite material that has
been used in the reactor environment forseveral years requiring an appropriate
disposal, hence minimise the total cost of
dismantling and decommissioning.
Accurate characterisation requires therepresentative samples be taken frommaterial to be chracterised. The spectrumof the radiation from the sample ismeasured and from this the constituentsand their activities are determined.
Assumingthat
thesampleisrepresentative of the entire component,
the total activity of the material
concerned per unit weight can be
deduced.
The graphite plugs which were used in
the vertical beam tube and irradiated forseveral years in the reactor environment
categorize as solid radioactive wastes,
these wastes are collected in polythinebags and safely stored in the waste

7Figure: Activated graphite material & corresponding γspectrum obtained from HPGe detector

Each graphite sample has been
weighed and measured on a calibrated
HPGe detector with 40% relative
efficiency coupled with Genie-2000
software for γ activity. In order tocalculate the activity of the graphite
sample, the energy and efficiency
calibration was done using a 152Eu
standard source and the spectrometer
was set to detect the dominant gamma
energy. The standard was placed on the
top of detector within the shielding
arrangement in the counting room of
HPRWMU. The counting time of thegraphite sample was 10,000 sec. The
sample was placed in the same position
as the standard source for efficiency
calibration. A same geometry waschosen for reference standard source for
efficiency calibration. An unirradiatedgraphite sample was also counted for10.000 sec.

of the measured radionuclides are: 6.3
Bq/g, 15.2 Bq/g, and 0.62 Bq/g for 60Co,
152Eu and 154Eu respectively. As
expected there is no measurable gamma
activity found for the unirradiatedgraphite sample. The result obtained inthis work could be used to estimate
radioactive waste generated duringdecommissioning activities and to
determine the appropriate methods for
dismantling and demolition for thisparticular nuclear facility.

Indonesia
Mr Dyah Sulistyani Rahayu,
Mr Suryantoro
National Nuclear Energy Agency(BATAN)

Integrity Test of the SNF Using
Sipping Test Method
INTRODUCTION
In the national policy an strategy on
radioactive waste management, SNF isclassifiedas
HLW(Goverment
Regulation No. 61 Year 2013
Radioactive
WasteManagement).
Indonesia follows open nuclear fuelcycle that the SNF will not be recycled.
CRWT as the centralized radioactive
waste management has responsibility to
manage the SNF from G. A. Siwabessy
Multi Purpose Reactor 30 MW.
CRWT has Transfer Channel Interim
Storage for Spent Fuel Installation (TCISSF). The facility was built in 1993 anddesigned by AEA Technology, United
Kingdom (UK-AEA). The main functionof TC-ISSF is to receive and store the
SNF and other irradiated materials. Mainpart of the TC-ISSF is water pond withdimension of 14 m in length, 5 m widthand depth of water 6.5 m. At the present
there are remaining of 245 elements of
SNF on the TC-ISFSF. The ISFSF has
the maximum storage capacity of 1448

Results
The activated graphite samples havebeen measured experimentally for
gamma activity using HPGe detector
coupled with Genie-2000 software. The
dominant gamma lines found in the
spectrum are: 1173.2 kev (0.99), 1332.5
kev (0.99), 121.8 kev (0.284), 344.3 kev(0.266), 1408.02 kev (0.21) and 723.3
kev (0.20), 1274.5 kev (0.355). The
identified respective radionuclides in theactivated graphite samples are 60Co,
152Eu and 154Eu respectively. From thestrong gamma lines the calculated
average specific activity concentrations
8

elements of SNF. Fuel element used in
GAS_MPR is Material Testing Reactor(MTR) fuel type with U-235 enrichmentof 19.75%, dispersed in Al matrix form
plate. Control fuel element is part of thereactor control rod that used to control
neutron flux. Based on GAS-MPR
technical spesification the fuel element
must be removed from reactor core after
reached 56% of burn up.
One of the requirements on radiation
safety and operation of a research reactor
is the absence of radionuclide releasefrom fission products to the environment.
Sipping test is one of non-destructive
testing technique for detection of thecladding by detection and identification
the presence of fission products in thewater such as Cs-137 and others.

Figure1: Transferring spent fuel from rack to sipping
test basket

2. The sipping tube is raised byrolling up the wire rope sling with
an electric motor until the top of
the test tube appears slightly above
the water surface.

1. THEORYSNF is a fuel element type Material
Testing Reactor (MTR) with a cladding
material is AlMg2. Being used in thecore of the reactor fuel elements
produce various fission products.
Fission products trapped in the matrix offuel elements can also separate out of thefuel matrix element. Noble gas group iseasy to separate such as Xe-138, Kr-87,
Kr-88, Kr-85m and Xe-135. The others
fission products and transuranics mayleak to the environmet since the isolationof the cladding has low integrity such as
defect, pitting or crack. Some of fission
products like Cs-137 pass through the
cladding by diffusion mechanism. The
sipping test is appropriarte method todetermine the integrity of SNF.

Figure2: Raising sipping test tube

3. Demineralised water is pumped
into the test tube through water
inlet port at the bottom of the tube.
It is estimated that 5 times test tube
volume of demineralised water isneeded to ensure that water inside
the tube has been fully replaced by
the demineralised water. Capacityof the test tube is 0.053 m3, so fivetimes of it is 0.265 m3. It takes
about 15.9 minutes to pumpdemineralised water using a pumpwith capacity of 1 m3/hour.

2. METHOD OF SIPPING TESTThe sipping test procedure:
1. SNF is transferred from its rackand placed into the basket of test
tube that located 2 m under water
surface.

98. Radioactivity analysis is conductedtodeterminetheactivity
concentration of Cs-137.
9. Previous step 5 to 8 are repeatedwith an interval of 4 hours.
10. Graph of Cs-137 activity
concentration versus time iscreated using the analysis dataobtained. Based on the graph, it
can be predicted whether the spent
fuel has leaked or not.
11. Water footage taken via pipeline
output and put into a test tube ofplastic bottles with a volume of
500 ml.
12. The activity of noble gases was
identified and measured using
gamma spectrometry device.

Figure3: Filling sipping test tube withdemineralised water

4. Water sample is taken from the test
tube for radioactivity background
analysis.

Figure4: Initial water sampling

5. Let the test tube with spent fuelinside for 4 hours.
6. Air is blown into the test tube tostir the water to be agitated.

3. RESULTThe sipping test method is one of thetechniques to find out possible leakage
on the fuel claddings, by detecting thepresence of fission product nuclides infuel elements soaking water. It has been
identified that 4 spent fuel element
releasing fission products with maximum
activity Cs 137 : 4,00 x 10-7 µCi/mL.
From these results indicate that there are
no leaks in the spent fuel detected.

Figure5: Stirring water inside sipping test tube

7. Water sample is taken from the test
tube for radioactivity analysis.

Figure6: Water Sampling

10

In Japan, AIST (National Institutesof Advanced Industrial Science andTechnology) keeps primary standard onradiation dose and dose rate quantities,
and some organizations have second orreference level standards of thesequantities with traceability to primary
standard.
Calibration system andmethods for radiation protection
instruments is standardized as a JapaneseIndustrial Standard (JIS) object (JIS Z
4511*1); Methods of calibration for
exposure meters, air kerma meters, airabsorbed dose meters and doseequivalent meters). Traceability system
in Japan is shown in Fig.1. Most ofcalibrationlaboratories
conduct
calibration of radiation protection
instruments according to the method of
JIS Z 4511 with standards traceable toprimary standard.

Japan
Mr Hiroyuki Murakami
Institute of Radiation Measurements
Prof Toshiso Kosako
University of Tokyo

A Measurement Quality
Assurance Activity for Radiation
Protection in Japan: Role of theInstitute of Radiation
Measurements
1. Measurement Quality Assurance ofRadiation Protection Instruments inJapan
Radiation measurements are very
important and necessary in and aroundany types of nuclear and radiation
facilities. In view of radiation protection
purposes at these nuclear related
facilities, reliability of the radiation
measurement results is of highly greatinterest. Especially after the accident at
TEPCO Fukushima daiichi NPP(Nuclear
Power Plant), public interest has become
higher than before and various types ofradiation protection instruments havebeen
commonly
used
for
the
measurement of radiation even in thegeneral environment.
However,reliability of radiation measuring
instruments (dose and dose rate meters,
radioactivity measuring apparatuses) is
not always sufficient. To get reliablemeasurement results, maintenance andcalibration system for keeping theperformance of these instruments ingood level (namely “measurement
quality assurance”) is strongly required.
Calibration and regular maintenancefor the instruments need 1) reference
standards with traceability to primary
standard and 2) engineers having good
capacities. Additionally, national and/orinternational quality assurance programs
are required to get higher reliability of
public and radiation workers.

Figure 1: Traceability System on Radiation Dose
Quantities in Japan

In Japan, more officially-certified
system called Japan Calibration Service
System(JCSS)
forcalibrationlaboratories (they satisfy ISO/IEC
17025*2) requirements) has been also
applied to get public-reliability on
measurement
resultsforvariousquantities. In the field of radiation
measurements, eight organizations areregistered as JCSS-fulfilled calibration
service laboratories, and three of them
11radiation knowledge throughout public,
such as seminars on radiation, etc., arealso conducted.
Some special activities of IRM aredescribed as follows;
1) Recently, needs for education andtrainings on radiation knowledge
are growing for very young
generations such as high school
students and thus IRM is now
strongly supporting to conduct
those programs by training highschool teachers.
2) Since 2009, IRM has planned an
information exchange meeting forexperts on current radiation
measurement technologies andcarried out it almost every year.
After TEPCO NPP accident inFukushima prefecture, various
typesofradiation
and/orradioactivity
measurement
instruments have been used atmany places and then the themes ofthe expert meeting are changed intosuch
relatedsubjects
asmeasurement
methods
atemergency situations, etc. Most
recently, IRM made an expert
meeting on agricultural foodproducts monitoring methods forcontamination area affected by
Fukushima daiichi NPP accident.
(See Photo 1)

have also been registered as aninternationally accredited calibration
service supplier (sufficient for ILACMRA system; International Laboratory
Accreditation Cooperation – Mutual
Recognition Arrangement). As one ofthree MRA registered calibration service
suppliers on radiation measurements inJapan, activities of the Institute ofRadiation Measurement are introducedin the following.
2. Activities of Institute of Radiation
Measurements*3)
In Japan, development of nuclearrelated technologies started in 1956, andsince then research and development
activities on radiation measurements had
also been mainly conducted at JAERI(Japan Atomic Energy ResearchInstitute; One of former organizations ofJapan Atomic Energy Agency: JAEA).
In 1980, JAERI constructed a large scale
facility for R&D on radiation
measurement for radiation protection
purposes, named Facility of Radiation
Standards; FRS, and intended to disperse
the productions of R&D activities onradiation protection measurements. The
Institute of Radiation Measurements(IRM) was established in the same year
(1980) to disseminate the fruitfulproductions of R&D activities of JAERI
and to build up reliability of radiation
protection
measurement
resultsthroughout social communities in Japan.
IRM is now contributing to society
through supplying reliable calibration
services of various types of radiation
protection instruments and conducting
relevant activities.
The business of IRM categorizes into
4 different activities, R&D activities onradiation measurements, calibration
servicesforradiation
protection
instruments,
measurementsonradioactive materials (including surveyof working environments) and trainingsof young generation on radiation
measurement techniques. Additionally,
many other activities to spread basic

Photo1: Expert Meeting on Radiation
Measurement Technologies(January, 2016 Tokyo)

123) IRM has been conducting an
investigation
measurement
program for widely contaminatedarea in and around Fukushima
prefecture since FY2013. This
investigation measurement (called
In-situ measurement) is made witha handy-type Ge detector and isuseful for analysis of distributionof radioactive materials after the
NPP accident. Photo 2 shows an
In-situ measurement work.

air absorbed dose meters and doseequivalent meters, Japan Standards
Association (2005) (In Japanese)
2) ISO/IEC 17025; General
requirements for the competence oftesting and calibration laboratories(2005)
3) Brochure of Institute of Radiation
Measurements (2013) (In Japanese)

Malaysia
Dr Mohd Abd Wahab Yusof

Malaysian Nuclear Agency

Borehole Disposal for
Disused Sealed Radioactive
Sources: Malaysian Initiative
Introduction
Sealed radioactive sources are widelyused by many industries, universities,
hospitals and research institutes in
Malaysia. All these practices have led to
the generation of radioactive waste, i.e.
disused sealed radioactive sources(DSRS), which some are intensely
radioactive. These DSRS have various
types of radionuclides which all need tobe management and disposed of safely
although some of them can be recycled
and reused. In Malaysia it is mandatory
by law to repatriate these DSRS to thecountry of origin. However, sometime
this cannot happen due to some reasonsi.e. the manufacturer or suppliers close
their business or bankrupt. If these
DSRS are not managed properly, they
can represent a significant hazard tohuman health and the environment.
Therefore the DSRS need to be sent to
Malaysian Nuclear Agency as National
Radioactive Waste Management Centre
for safe keeping after getting approval
from the regulatory body.

Photo2: In-situ Measurement in environment(Fukushima Prefecture)4) In Japan, there are 4 different
individual monitoring service
companies
that
cover
approximately 5 hundred thousand
radiation workers. In order toverify that the uncertainty of eachservice will be within thepermissible level, IRM is playing acentral role of individual dosimetry
inter-comparison program; blind
irradiation and report of themeasurement evaluation results toeach companies.
IRM is a small organization(about 50 personnel) but, as
introduced above, is now playing
an important role for securing thereliability (quality assurance) on
radiation measurements activities
in Japan.
References:
1) JIS Z 4511; Methods of calibrationfor exposure meters, air kerma meters,13

Continued storage perpetuates
the risk and places burden on
future generations
 Concerns over safety and security.
 High chances of accidents tooccur.
Borehole Disposal Concept
The Borehole Disposal Concept (BDC)
has two main safety functions – it allows
waste packages to be placed deep below
the ground where humans cannot
interfere with them, and it provides
access to an environment that is stable
and capable of keeping the waste
packages intact for many thousands ofyears. Physically, a BDF is about aquarter of a metre diameter and morethan 30metres to 300 metres deep. It will
be drilled and constructed using standard
techniques developed for deep water
abstraction, oil exploration etc.
The borehole will usually be lined withmild steel or high density polyethylene
tubing with concrete pumped into thegap between the lining and the host rock.
The purpose of the lining is to aid the
packageemplacement
process,
especially by supporting the borehole
wall (see Figure 2). At the bottom of theborehole is a concrete plug. Once theconcrete has set, a fresh batch ofconcrete is placed in the bottom of thehole and the first disposal package islowered into this so that it sinks throughthe fresh concrete and stands on theconcrete plug. When this batch ofconcrete is set, the top of the concrete
forms the platform for the next disposal
package. The process is repeated until
the required number of packages has
been emplaced. The topmost disposal
package will always be at least 30 metres
below ground level.
With all the packages in place, thesection of borehole lining above thetopmost package is removed. A steel
deflection plate is placed above thetopmost package to help prevent
inadvertent drilling into the disposalzone and the (now unlined) top of the

Figure1: Disused sealed radioactive sources

Storage in a secure facility can beconsidered as an adequate final
management
optionforsources
containing quantities of short-lived
radionuclides, which decay to harmlesslevels and the radioactivity below thanregulatory control within a few years.
However, for most other sources asuitable disposal option is required.
Near-surface disposal facility is notsuitable for the source constitutes a high,localised concentration. Deep geologicaldisposal offers the highest level ofisolation available within disposal
concepts but such facilities are underconsideration for the disposal of spent
nuclear fuel, high level waste andintermediate level waste in a number ofcountries. However, they are expensive
to develop and only viable for countries
withextensive
nuclearpower
programmes.
Therefore
increasing
attention has been given to the disposalof DSRS with long lived radionuclides
and very high activity sources inborehole facilities with a view toproviding a safe and cost effective
disposal option for limited amounts.
It is the right time for Malaysia to
consider the borehole disposal as theultimate solution to the increasing
number of DSRS stored in the current
facility. This is because: Many DSRS cannot be returnedto the manufacturer.
 Inadequate storage of DSRS
presents

14

At present a Malaysian Nuclear
Agency is carrying out a feasibility study
to develop BDF. Desk study on geology,hydrology (surface and underground),geochemistry,
geomorphology,
meteorology and seismic has been done.
The feasibility study also includes
studying several potential sites anddetermining whether the potential site issuitable for accommodating a BDC. The
suitable site will only be considered for
future work if provided the safetyassessment demonstrates of its feasibility,
otherwise the facility will be sited at a
yet to be determined alternative location.

borehole is filled with concrete to within
2 metres of the surface. The final few
metres are filled with native soil. This
will make the borehole difficult to locate
without specialised equipment. Figure 1
shows the schematic layout of a borehole
disposal facility.

DSRS Inventory, Characterisation
and Conditioning
Currently, there are only 4 DSRS in
Category 1 and Category 2, and morethan 12,000 DSRS are in Category 3 to 5
in Malaysia. The current priority is todispose of the Category 3 to 5 DSRS
under the borehole disposal concept.
With regard to the Category 3 to 5 DSRS,
the sources will be identified from thecurrent inventory, characterized andverified (Figures 3 and 4). The sources
are then dismantled from it shielding
container and emplaced into a stainlesssteel capsule (See Figure 5). The capsulewill be remotely welded, leak tested andplaced into a shielded concrete-filled
200-L steel drum. The drum with welded
capsule will be transferred to a storage
facility. Emplacement into a disposal
container and welding will be done in
the next stage of the process. All thesework will be done within a properly
shielded and controlled work area (SeeFigure 6).

Figure2: Schematic layout of a borehole disposal
concept (Ref 1)

Why is Malaysia interested withborehole disposal concept?
Malaysia is interested to develop
borehole disposal concept. A study toknow the suitability of borehole disposal
concept in Malaysia has been done. It
was found that the advantages of theborehole disposal as the following: Adjustable depth – the depth canbe adjusted between 30 metres to
300 metres depend on the numberof disposal canister to be disposedof; Safety – comply with the “defence
in depth” concept where there are 8tiers defence which include
radioactive double encapsulated
stainless steel, stainless steel 316L
capsule, concrete barrier, stainlesssteel 316L disposal canister,
concrete backfill, steel or HDPE
casing, concrete disturbed zone
backfill and geosphere/host rock.
 Security - robust to human
intrusion Small footprint Simple, easily available technology
 Small volume excavation

15

Figure3: Sources Characterization

Reference
1.
International Atomic Energy Agency,
Borehole Disposal Facilities for Radioactive
Waste, IAEA Safety Standards Series No. SSG1, IAEA Vienna (2009)

ThailandMs Nanthavan Ya-anantThailand Institute of Nuclear
Technology (TINT)
Ms Pennapa Kanchana
Office of Atoms for Peace

Figure4: Sources Characterization

Safety and Security during
Transport ofRadioactive Materials andRadioactive Waste
The Regional Workshop on Safety andSecurity during Transportation ofRadioactive Material was conducted in
Bangkok, hosted by International
Atomic Energy Agency (IAEA) and
Office of Atoms for Peace (OAP),
during 22-26 June 2015. The purpose of
the workshop is to present and discuss
experiences and best practices in thedevelopment and implementation ofsafety requirements and security
recommendations on the transport ofradioactive material, national andinternational cooperation and preventionof delay/denial of shipment ofradioactive material.

Figure5: Conditioning of DSRS done by Nuclear
Malaysia Staff

The expected outputs of the workshop
are:
1. clear understanding of theobjectives of the IAEA safety
requirements
andsecurity
recommendations for transport ofradioactive material;
2. general
understanding
andappreciation of the similarities
and differences between safety
requirements
andsecurity
recommendations
duringtransport of radioactive material;

Figure6: Work Area for Conditioning of DSRSConclusion
This project is expected to develop a
Borehole Disposal Facility for disusedsealed radioactive sources. The current
and future generation will benefit fromthis project. The success of the project
would set as an example and experiencecan be shared with other countries as
borehole disposal is not available
anywhere else in the world.

16

3.

4.

The technical visit was conducted to visit
a company which carried on an activityof transportation of radioactive sources
for gamma radiography , as shown in
Figure 2.

general appreciation of theinterface between transport safety
and security and potential
impacts of this interface on theefforts
tostrengthen
thecompliance assurance regime forthe safe transport of radioactive
material
presentation of experiences andgood practice in the development
andimplementationofrecommendations, national andinternational cooperation andprevention of delay/denial of
shipment of radioactive materialtransport

There were 25 participants from theMember States and Entities in Asia andthe Pacific, participating in the regionalTC
project
RAS/9/067,
namely:Afghanistan,
Australia,
Bahrain,
Bangladesh, Cambodia, China, Fiji,
Indonesia, Islamic Republic of Iran, Iraq,
Israel, Jordan, Kuwait, Lao People’s
Democratic
Republic,
Lebanon,
Malaysia, Marshall Islands, Mongolia,
Myanmar, Nepal, Oman, Pakistan, Palau,
Philippines, Qatar, Saudi Arabia, Sri
Lanka, Syrian Arab Republic, Territories
under the Jurisdiction of the Palestinian
Authority, Thailand, Vietnam andYemen, as shown in Figure 1.

Figure2: Transportation of radioactive sources
for gamma radiography

Radiation Safety during Transportation

CONTAMINATION LIMITS
The
overriding
requirement
concerning the presence of radioactive
contamination is ALARA
Contamination
limitsareestablished for:-External surfaces of all packages-Internal surfaces of tanks, overpacks, IBCs and freight containers
Non-fixed contamination on theexternal surfaces of packages and on
the internal and external surfaces ofoverpacks, freight containers, tanks
and intermediate bulk containers shall

Figure1: Participants from Member States

The workshop consists of presentations,discussions, and practical sessions on
safety and security during transport ofradioactive material including national
and international cooperation andprevention of delay/denial of shipment.
17

be kept as low as practicable and shallnot exceed the following limits -For β, ϒ and low toxicity α
emitters: 4 Bq/cm2
For all other α emitters: 0.4
Bq/cm2

Figure3: Transportation of Radioactive
Materials and Radioactive Waste in Thailand

TransportationofRadioactive
Materials and Radioactive Waste in
ThailandThailand Institute of Nuclear
Technology(TINT)
hasthetransportation service for radioisotopes for medicine and radioactive
waste in Thailand as shown in Figure 3.

18

The FNCA RS&RWM Project Leaders
Australia

Bangladesh

Ms Lynn TAN
Project Transition Leader
Waste Projects and Strategic
Planning
Australian Nuclear Science andTechnology Organisation

Dr Moinul Islam
Chief Scientific Officer,
Health Physics & Radioactive
Waste Management Unit,
Atomic Energy Research
Establishment (AERE), Bangladish
Atomic Energy
Commission (BAEC)

China

Indonesia

South Korea

Dr Chang Lak Kim

Malaysia

Dr Mohd Abdul Wahab
Yusof
Director,
Waste and Environmental
Technology Division,
Malaysian Nuclear Agency(Nuclear Malaysia)

Mongolia

Dr Zhang Jintao

Director of Center forRadioactive Waste Technology,
National Nuclear Energy Agency
of Indonesia (BATAN)

Ms Oyuntulkhuur
Navaangalsan
Head,
Radiation Safety Council
Nuclear Energy Commission

Deputy Director General,
Department of Safety, Protection
and Quality,China National
Nuclear Corporation (CNNC)

Mr Suryantoro

Director,
Radioactive Waste
Management Policy Office,
Korea Radioactive Waste
Management Corporation(KRMC)

Philippines

Ms Maria Visitacion B.
PalattaoHead, Regulations andStandards Development
Section, Philippine Nuclear
Research Institute (PNRI)

Ms Nanthavan Ya‐anantJapan

Kazakhstan

Prof Toshiso Kosako

ThailandProfessor,
Nuclear Professional School,
Graduate School of Engineering,
The University of Tokyo

Mr Yevgeniy TUR
Technical and Economic
Assessment Department
National Nuclear Center of the
Republic of Kazakhstan

19

VietnamHead, Radioactive Waste
Management Section,
Radioactive Waste
Management Center,
Thailand Institute of Nuclear
Technology (TINT)
Dr PHAM Quang Minh
Director, Department for
Planning and R&D
Management,
Vietnam Atomic Energy
Institute (VINATOM)

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