FNCA
Forum for Nuclear Cooperation in Asia

RS&RWM
Newsletter No.14
Topics from FNCA Countires
March 2018
Issued & Edited by FNCA Secretariat

Indonesia
Mr Mr. Moch Romli
Mr Sucipta
National Nuclear Energy Agency(BATAN)

Hazard and Operability Study for Designof Near Surface Disposal in Serpong Site
Introduction
In order to increase capacity building in nuclear
technology, especially in radioactive waste
management, BATAN has a near surface
disposal design for the Serpong site. It is
designed to accommodate low level radioactive
waste currently stored in the interim storage at
the Center for Radioactive Waste Technology BATAN.
The very basic principle of disposal is that
disposal facilities are placed, designed,
constructed, operated, closed anddecommissioned in such a way that theirworkers, publics, and the environment areprotected from radiological hazards. In
preparing the design of Near Surface Disposal(NSD), the dimensions of waste processed andthe dimensions of building its own NSD
facilities have to be prepared and become thebasic data that must be calculated. The size,
weight, and amount of waste to be placed in the
NSD should be estimated, it is necessary tomaintain the condition of the packaging and theplacement of the waste.
Near Surface Disposal building foundations
should be able to support the overall load. The
disposal site to be built is not only used as amere waste disposal, but needs to be equippedfor various things, such as the addition ofbuildings to complete the disposal operation.
The layout and addition of buildings vary fromone disposal facility to other disposal facilities,
depending on need. Most importantly, thesupporting facilities within the disposal area
should be designed in such a way as to facilitate
the decontamination of an accident resulting incontamination within the facility's disposal area.
Roads connecting waste receiving areas andother supporting areas with disposal areas
should be designed separately so that disposal
activity traffic does not affect disposal structuresthat have been filled with radioactive waste, andthe presence of supporting facilities does notinterfere with disposal activity traffic.
Geomorphologically, Serpong site relatively
meets the requirements with hilly and nonsloped contours, the slope is small and theintensity of the geomorphology process is also
small [4]. Although in this aspect the existingsite has met the criteria, but optimization is still
done by placing disposal facilities at locations
as far as possible with rivers, slopes, and zones
that have large erosion potential.

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The construction of a demo disposal facility isplanned for an area of 1197.16 m2 with a size of
34.6 m x 34.6 m. Engineered vault shape withreinforced concrete vault walls with 0.6 m wall
thickness. Vault is built with a depth of 2 m andthe foundation of the building is located 4 m
above the highest groundwater level. Vault isdivided into 2 (twin) compartments separated
with 0.6 m thick reinforced concrete walls. The
volume of each vault space is 594 m3 with alength of 18 m, width of 6 m and a depth of 5.5
m. Right compartment for disposal of waste
package in 200 liters drum with high drum
dimension 85 cm and diameter 60 cm withweight ± 0,486 ton. While the left compartment
for the waste package in a 950 liter concrete
shell with dimensions of height 1.4 m and adiameter of 1.3 m with a weight of ± 6.4 tons.
For the left compartment, 200 liters drums arearranged in 15-lane, 15 row, and 3 stack
modules, so that in one module it can hold 675
200-liter drums. For the left compartment the
950 liter concrete shell will be arranged into 6
lanes, 6 rows, and 2 stacks, so that in one
module can accommodate 72 concrete shells.
Methodology
This study is an analytical descriptive study toidentify hazards on the design for NSD. The
analysis is carried out on any potential failuresof the disposal system and its supporters, andprovides descriptive descriptions of the factors
leading to such failures and their impact onsafety, working areas, the environment
surrounding NSD facilities, and financial losses.
In this study, an analysis of the design ofdisposal system components consisting of thedisposal system itself (cover, vault, drainage,
waste package, backfill), site characteristics, andaccess control of disposal facilities.

Figure 1. Cross-Sectional View for Design of NSD

1S, 2S : Concrete Shell Storage
1D, 2D : Drum Shell Storage
3 : Laboratorium
4 : Utilities Building

5 : Maintenance Building
6 : Interim Storage
7 : Main Gate (In – Out)
8 : Restricted Area BorderFigure 2. Layout for Design of NSD

HAZOPS is chosen because that it is asystematic, thorough, and complete method, andcan be used in the design stage of an installationor production process. The design stage is theright time to do the HAZOPS study. If theresults of the HAZOPS study recommend
corrective action against the design, then it will
not cost more than later when the facility orprocess has been running.
Results and Discussion
The study node of the radiation safety risk
analysis is the barriers in the disposal system.
Barriers are nodes of study because they are thedeterminants of the reliability of the disposal
system to support the radioactive waste stored
therein, either as shielding for radiation

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exposure or containment of the possible releaseof radionuclides to the environment.
Generally, the barriers are divided into 4 (four)
disposal components i.e. disposal engineeringsystem, waste package, site, and access control.
The disposal system consists of cover, vault(including bottom cover), drainage, backfill
material and buffer used. As for the waste
package, the integrity of the waste matrix is thedeterminant of the reliability of the waste
package as a supporter. For site risk assessment,
various site-related risks will be assessed
including geomorphology, lithostratigraphy,
geological structure, engineering geology,hydrogeology, and potential natural disasters.
Whereas access control has a potential failure ofaccess should not, whether it is access without
intrusion, or access with intrusion. The
parameters present at this study node aredetermined at the design stage, so the results canto be a recommendation for the design.
Tabel 1. Study Node for Design of Near Surface Disposal

No.

Study NodeParameter

1.

Cover

Permeability

2.

Backfill

Permeability

3.

Drainage

Flow Rate

4.

Waste Matrix

Integrity (compressive
strength,density,leachate test)

5.

Vault (include Radiation Dosebottom cover)
Permeability

6.

Access Control

7.

Site
(Natural Permeability, distance ofBarrier)
facility to surface water
body,
soil
bearing
capacity

Rate,

Incongruity

1. CoverCover is built after the operation phase of thedisposal facility. Cover must satisfy 3 (three)design criteria, namely impermeability,
reliability, and protection. In the impermeability
criterion, the amount of rainwater that enters thecover and reaches the waste package must bevery low to avoid leaching and radionuclide
migration. The criterion of reliability is that theimpermeability of the cover must be maintained
until the period of institutional control or wasteends for approximately 300 years. The material
to be used should be selected to ensure its
ability over a long period of time. While thecriteria of protection require the cover to beresistant to external forces such as erosion,
chemical processes, and organisms. Of all thesecriteria, permeability becomes a parameter that
determines the reliability of the cover. For thatcover must be made thick enough and made ofmaterial that resistant from external force. If thecover fails to maintain its performance,
significant water infiltration will occur and will
impact on the performance of the waste
packaging that will be the direct cause of therelease of radionuclides from the waste
packaging.
Serpong Nuclear Area is an area that has highrainfall and tropical climate so that proper cover
design should be considered for its function as adrain and rain water drainer can run well so thatrain water does not contact with radioactive
waste package. The recommended cover design
is as follows: The cover component is a multi-layered
cover, consisting of an uptake layer that alsofunction as a biological protective layer ofvegetation with soil, rocks, and sand, then alayer of low permeability that can be clay or
bentonite compacted, primary and secondarywith membrane layers placed between 2(two) drainage layers;

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 The thickness of each layer depends on thetype of radioactive waste. Since the locationof the potential disposal facility has highrainfall, the thickness of the drainage layer
should be thick enough with the dome design
having a slope of 2o – 3o; The drainage system is a part of the cover
design comprising primary and secondary
layers so it must be carefully calculated in
the overall design to ensure the function ofradioactive waste insulation.
The recommended cover design is the result ofcover performance analysis done by using ElCabril cover design.
2. Backfill
Backfill or filler material is engineered barrier
added to sub compartment after fully charged.
Backfill fills the gap between the waste package
and the layer between the waste package and thelayer above it. The main function of the backfill
is as a high permeability layer that passes water
to immediately descend to the bottom of thedisposal if there is water infiltration (either
surface water or ground water) into the vault
thus reducing water contact with packaging andwaste matrix. If the backfill fails to perform its
function, then the chance of water contact withthe waste package will be greater, it means the
chance of damage to the waste package will also
be greater.
Anticipation of potential failure of backfill
material can be done by laboratory scale
experiments on the type of material that has
high permeability with the existing site
environment. According to the results of thestudy, a mixture of bentonite and quartz sand(more quartz sand composition) was used.
3. Drainage System
The drainage system in the NSD facility serves
to drain the infiltrated water into the facility to

the control pool so as not to accumulate in thefacility. If water accumulates in the facility, it
will increase the moisture inside the facility thatcan enlarge the waste pack degradation factor.
And the worst possibility can cause flooding in
the facility that inundated the waste package.
Risk control measures that can be done toprevent the failure of this drainage system is byengineering drainage-shaped gutter with a slopeof 1%.
4. Waste Matrix
As the waste packet node representation, waste
matrix is used which can be measured its
characteristics with its own integrity.
Radioactive waste that has been treated with amixture of cement and sand, both in 200 liters
drums and 950 liters of concrete shell, always
tests quality to determine the integrity of thewaste matrix. The integrity of the waste matrix
can be seen from the results of compressive
strength test, density measurement, and leachate
test. The results of the compressive strength test
and leachate test show how strong the waste
matrix from the risk of breakage and how
resistant to the waste matrix will not decompose
which may lead to the release of radionuclides
from the waste package. As a recommendation,
it is advisable that the waste package to be
stored in the NSD facility is the waste treatment
package for the last 10 years. This is toanticipate the degradation of waste packages
that have occurred during temporary storage. If
not possible, then the package of waste to be
stored viewed the condition of packaging, if it
has experienced degradation should be overpackaging.
5. Vault (include Bottom Cover)Vault function as the primary supporting of thedisposal system. In addition it function as amajor supporter of the disposal system, i.e. as aradiation barrier from radioactive waste and as abarrier (vault and bottom cover) to the release ofradionuclides from a packet of radioactive waste

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to the environment. If this node fails, it will be
indicated by 2 (two) major impacts, namely theincrease of radiation exposure and the release ofradionuclides. For a radiation exposure barrier/shielding, the selection of materials that havesufficient half-value layers to withstand theradiation from the encoded radiation source.
While its function as a containment is highly
dependent on the quality and homogeneity of its
constituent material in order to survive thehydrological/ hydrogeological andclimatological environment around the NSD
facility tread.
As an effort to control the risk of exposure toradiation outside Vault, in the pre-design stage it
is necessary to know in advance the inventory ofradioactive waste to be stored in the NSD
facility, be it radionuclide, activity, radiation
exposure, and amount of radioactive waste to be
stored in the facility. If this inventory is known,
then the design for the vault can be determined,
such as material with sufficient half-thickness aswell as the thickness to be satisfied. For therecommended vault design is K350 reinforced
concrete 60 cm thick. This concrete other thanas a radiation shielding, also serves as areinforcing structure in order to optimize thedisposal placement in the Serpong Nuclear Areasite. As an early warning effort, routine
monitoring of the pool can be monitored to
anticipate releases of radionuclides in order not
to release to the environment. As a supporter ofradionuclide release, and taking into account thesite conditions, rainfall, relatively shallow
groundwater surface and the potential for
movement of the soil, a below grade vault
design is recommended.
6. Control Access
The main function of access control is to restrict
unauthorized parties from entering the NSD
facility. If this access control fails, there may be
intrusions that cause direct or indirect disposal
facility damage, as well as acceptance ofradiation doses that should not be accepted. The
access control function shall be maintained
throughout the operating period of the facility
until the end of the active supervision period(approximately for 300 years).
Risk control measures that can be taken toprevent the failure of access control startingfrom engineering by placing NSD facilities in alimited environment equipped with security
posts, fences, doors, alarms and monitoring
cameras. For NSD facility candidates to be built,
it has fulfilled several requirements as it isplaced in Serpong Nuclear Area (SNA) at
PUSPIPTEK, which must at least have to pass 3(three) layers of guard post before reaching
NSD facility. In addition, administrative
arrangements (standard procedure) need to be
set up for access control and control
arrangements in case of nonconformities.
7. Site Characteristic
For Site node, there are 3 (three) main
parameters that can be failure factor. First, whenviewed from the permeability parameter, SNA
site has a higher value than the specified
criterion. This will have an impact on thepotential for water infiltration to a larger NSD
facility. The second, the NSD facility distanceto surface water bodies (rivers), the location of
the NSD facility is only about 150 meters fromthe Cisalak River. Not in accordance with theideal criteria that require the location of NSD
facilities to the surface water body of more than
500 meters. This will affect the resistance of
natural barrier that is smaller than expected.
This means that if there is a release ofradionuclides, then the possibility ofradionuclides that reach the surface water bodies
will be greater. The third is the carrying
capacity of the ground. The site where the NSD
is built is an area that has rock hardness and lesscompactness, and low homogeneity. With thecondition of such areas, if forced to build
facilities disposal demo in the area without

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engineering it will result in unstable facilities
and will experience a rapid decline.
To anticipate the risk due to some characteristic
mismatches of the site against the ideal criterion,
strengthening is done in the engineered barrier
disposal system. The engineered barrier of vault
is selected with a low permeability material ofreinforced concrete K350 with a thickness of 60
cm to ensure that no radionuclide release occurs.
To lower the probability of failure of the facility
due to being built in areas with low soil carrying
capacity, the demos disposal facility will bebuilt in the vadose zone by using deep
foundation.
Conclusion The NSD Facility has potential hazards,
particularly hazards to radiation safety as it
may cause environmental pollution due to therelease of radionuclides from waste packages
stored in the facility. The determinant of
releasing the radionuclides to theenvironment is the reliability of the existing
barrier in the disposal facility. To identify thehazards of failure of each barrier, the
HAZOPS method is used; Hazards arising from NSD facilities,
particularly radiation hazards associated withenvironmental pollution and radiation
exposure, are caused by design errors andhuman intrusions that may cause deviations
from the barrier function. Some parametersof the barrier design that can be used to
calculate and estimate risks during the period
of operation and active post-closure
surveillance are material permeability, theintegrity of the waste package, the drainage
flow rate, the carrying capacity of the soil,
and the nonconformity in access control; The use of HAZOPS studies in this study inaddition to estimating future risks, may also
be used to provide appropriate design,material, and safeguards recommendations
for disposal facilities.

References
Susilowati, D. Preparation of the DemonstrationPlant of Disposal Layout Design Component in
Serpong Nuclear Area. Technical Document"Concept of Demo Disposal Facility in Serpong
Nuclear Area and Characterization of Selected
Site Disposal of Radioactive Waste" (pp. 20-29)
Tangerang Selatan: Center for Radioactive
Waste Technology – BATAN (2013).
Sucipta, Demo Disposal Placement Optimization in
Geological Environment at Serpong Nuclear
Area,
JournalofRadioactive
WasteManagement Technology Vol. 2, CRWTBATAN (2013).
Wachyudi, Y. Hazard Identification, Analysis, andRisk Control in the Design Phase of Oil & Gas
Production Process on Floating Production
Storage & Offloading (FPSO) Ship for Petronas
Bukit Tua Project 2010. Depok: University ofIndonesia. (2010).
Colling, D. A. Industrial Safety : Management andTechnology. New Jersey: Prentice Hall. (1990).

Annex Tabel : HAZOPS for Design of Near Surface Disposal on Serpong Nuclear Area

No.
1.

Node
Cover

Parameter
Permeability

Guide Word

Deviation

High

High
permeability

2.

Backfill

Permeability

low

Low
permeability

3.

Drainage

Flow rate

nothing

There is noflowrate

4.

Waste Matrix

Integritylower

Low integrity

Higher

Dose rate
higher thandoseconstraint

Dose rate

5.

6.

7.

Vault

Control Access

Site

Cause
incompatibility ofdesign andmaterial selectionl
selection of
inappropriatematerial
composition
the gutter tilt is
not perfectimproperimmobilization
process
incompatibility ofradioactive waste
exposure withmaterial type andwall thickness ofvault

Permeability

High

High
permeability

material selection
error vault,
erosion, porous

Access control
incompatibility

occourthere isundue accessinsufficient access
control

Permeability

HigherHigherpermeability

availability ofarea

Distance
facility to water
body
Soil bearing
capacity

less

less

Distance
facility towater body< 500 mThe carrying
capacity ofthe soil is not
ideal

ImpactRainfall infiltration
increase theprobability ofcontacting the waste
package with water
Accumulation of
infiltrated water

Risk Control

Recommendation

El-Cabril design adapted

Monitoring of rainfall
infiltration during
active surveillanceselection of appropriatematerial composition
Gutter with slope 1%

radionuclides areeasy to release

cover, vault, drainage,
control pool

Waste AcceptanceCriteria

Individual dose
increaseselection and calculationof material thickness
appropriate for vault
material, WAC

vault using K350
reinforced concrete 60
cm thick

Water infiltration tofacility

selection of vault materials
according to sitecharacteristics ingeological andclimatological aspects

vault using K350
reinforced concrete 60
cm thick, periodic
monitoring of control
pool as an early
warning

intrusion, receptionof radiation doses
should not be
There is infiltration
potential andradionuclide release

post guard, fence, alarm,
cctv, lock settings
vault with K350 concretewith 60 cm thickness

availability ofarea

natural barrier weakvault with K350 concretewith 60 cm thickness

less rockhardness, lesscompactness,
homogeneity

unstable facility,
decreased fasterdeep foundation of thevadose zone

6

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has already completed the site screening in
2011.

Malaysia
Dr Norasalwa Bt. Zakaria
Malaysian Nuclear Agency

Extending the Borehole Disposal
Experience for the Development of the
National Low Level Repository Program
1.0 Introduction
National programme on the development of nearsurface repository in Malaysia was reviewed in2016 and the program was strengthened by
identifying gaps in the existing program.
Among the gaps identified are the lack ofexperience from the significant number of new
staffs and the lack definition of the boundary
conditions of the repository. Thus, in order tomove forward, the project has identified
priorities areas namely:i. Priority 1 – Inventory. The full inventory of
LLW (and possibly ILW) to be disposed of in
the national repository need to be defined aswell as the final waste form.
ii. Priority 2 - Safety assessment, Safety case andRepository design. For a start, Nuclear
Malaysia is planning to undertake a genericsafety assessment, in order to initiate thetraining programme and developing tacitknowledge in the respective fields.
iii. Priority 3 - Stakeholder engagement andcommunication. Normally, this involves a longprocess to talk to the public, to enhance
understanding and to obtain acceptance.
However, this process need to start as early as
possible, using all available medium.
iv. Priority 4 – Siting campaign. Full set of thesiting process is activated that includes
selection of potential site, candidate site and sitecharacterization. The siting process begins withthe site screening, of which Nuclear Malaysia

At this point of writing, Nuclear Malaysia ispreparing to begin Priority 1 and Priority 2
concurrently, taking advantage the experiencefrom the Borehole Disposal System Project
which specifically focusing on the disposal of
Category 3 to Category 5 disused sealed
radioactive sources.
2.0 Experience from the Borehole Disposal
System Project
2.1 What is a Borehole Disposal System
Malaysia is planning to construct a Borehole
Disposal System (BDS) for the disposal of
Disused Sealed Radioactive Sources. It is a
specially engineered drilled boreholes, with anengineered barrier system, based on the use ofstainless steel capsules and buries in a boreholeat depths greater than depth 30 meters of thatnear surface disposal facility. The BDS falls
between the two well-established options ofdisposal in near surface facilities and disposal ingeological facilities.
2.2 Project Activities
As a type of a radioactive waste disposal facility,
the overall development and implementation of
a BDS project constitutes five phases which are
the planning phase, design phase, construction
phase, operational phase and post closure phaseas shown in Figure 1.

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the borehole. Right after closure, the BDF enters
the post closure phase and activities dedicatedduring this phase are institutional control andeventual release from institutional control.
2.3 Site selection and characterization
The siting guidelines for the BDF system ismuch less stringent than those of the nearsurface and deep geological disposal system as
the BDF design provides for the radiologicalcontainment in a wide variety of geologicalenvironment.
The site characterization activities wereconducted for 2 years from 2014 until 2016. The
main objectives of the site characterization
activities were to observe, inventory, anddocument the presence, type, location, andrelative amount of geological, hydrogeological,
geotechnical properties and hydrochemistry.
Geological mapping has been carried-out withtechnical assistance of Minerals and Geoscience
Department Malaysia.
Figure 1: Implementation activities for the borehole disposal
facility

During the planning phase, project activities thatare being undertaken involve updating thesource inventory, identifying design
requirements and performing a desktop
preliminary study. Activities performed at the
pre-disposal phase includes site selection, sitecharacterization, waste acceptance criteria
established, waste conditioning that involve a
series of source retrieval, encapsulation andcontainerization, site-specific design of the BDS,
safety assessment and safety case, license
application, manufacturing and qualification ofequipment and last but not least, is theconstruction of the Borehole facility.
Operational phase or also referred to as thedisposal phase constitutes disposal activities liketransport and transfer of filled and sealed
disposal containers to the borehole, associatedbackfill of the borehole and finally closure of

Two exploratory boreholes were constructed todetermine the actual thickness of the residual
soil cover, the actual depth of the contact among
the metasedimentary rock underneath as well as
to delineate any subsurface geological structuralfeatures that intersect the proposed disposal
borehole. The depth of the exploratory
boreholes exceeded 200 m below the existing
ground level. Core logging, geomechanic
properties (such as mineral density, bulk density,
bearing capacity), hydrogeological testing (suchas permeability/hydraulic conductivity,
groundwater flow, recharge/discharge) andgroundwater sampling for geochemistry testing
as well as down hole geophysical logging havebeen carried out during the exploratory
boreholes drilling and after the completion of
the drilling work.
The full set of site characterization study covers
the assessment of local geology, regional
geology, seismic analysis, topography,

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hydrology, surface erosion, hydrogeology andhydrochemistry. Most of the data that are
obtained from the site characterization are used
to construct the conceptual models on
geological, hydrogeological and geochemical
conceptual models. This geological conceptualmodel demonstrates the geological condition of
the proposed site including the geomorphologic,
bedrock condition (lithological, deformation
zone, fractured, others structural of thesoil/rock). It provides the geometricalframework and the geoscientific descriptions
necessary for development of the rockmechanics models, hydrogeological models and(hydro) geochemical models, which are required
for simulating repository behaviour on the shortand on the long term, i.e. for the design andconstruction of the repository and for thedemonstration of safety.

Figure 2: Conceptual model of radionuclide release and migration
processes in BDS

2.4 Post-closure safety assessment and the
Safety Case
The post-closure safety assessment is following
the Long Term Safety Assessment
Methodologies for Near Surface Radioactive
Waste Disposal Facilities approach, or in short,the ISAM Safety Assessment Approach. The
approach consists of the following key steps:-

The specification of the assessment context;
The description of the disposal systems;
The development and justification of scenarios;
The formulation and implementation of models;Quantitative analysis was derived using
AMBER calculational code. In addition,
deterministic approach was employed to assess
the sensitivity of parameters used in theconceptual model.

Figure 3: Example of calculated annual dose profile in the postclosure safety assessment

2.5 Project Management
In order to implement the project structurally
and efficiently, a project management team wasformed that comprises 7 working groups,
namely site selection and characterization group,
borehole construction and implementation group,
safety assessment and safety case group, waste
characterization group, waste conditioning
group, licensing and research and developmentgroup. The project management structure isshown in Figure 4.

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breadth of a repository project. Training andhuman capital development should be properly
planned and been given an emphasis. Recent
training under the Nuclear Researchers
Exchange Program, under the purview of
MEXT, Japan were carried out on researches
related to repository development. The name of
the trainee, title of the research project and theyear conducted are given below:

Figure 4 : Project Management Structure

Altogether, there are 30 core personnel in thisproject comprising of officers and supporting
personnel from several divisions and units in
Nuklear Malaysia. In addition to strengthening
the technical capacity and capabilities,
collaboration is made between Nuklear
Malaysia (Ministry of Science, Technology andInnovation) and the Ministry of Environment
and Natural Resources, where 2 professional
geologists from the Mineral and Geological
Department, were attached in this project in thesite selection and characterization group.
3.0 Extending the BDS Experience for the
Near Surface Repository Project
A near surface repository is a large scale projectencompasses elements of siting, safety case,design, stakeholders communication and etc.
For such project that involves long timeframe, it
requires sustainable supporting system inter alia
strong project management, efficient knowledge
management, and properly planned training ofstaffs Clearly, the experience gathered during
the borehole project is valuable in partly
accelerating the technical aspects know-how inparticularly with respect to site characterizationand post-closure safety assessment.
The training required to support the repository
development should cover the overall depth andi. Azmi Ibrahim – Investigation of
Biosphere Model in Safety Assessmentof Radioactive waste disposal,
University of Fukui, Japan. Year 2016
ii. Kang Wee Siang – A dynamic model
approach to simulate current practice ofradioactive waste management in
Malaysia, University of Fukui, Japan.
Year 2017

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Dr M. Moinul Islam1
Dr Kandoker Asaduzzaman2
Chief Scientific Officer1, Principal
Scientific Officer2 HPRWMU, INST,
AERE, Bangladesh Atomic Energy
Commission

Radio-chemical Characteristics of Liquid
Radioactive Wastes
Pre-treatment of the liquid waste is the initial
step in waste management that occurs after
waste generation. It consists of collection,
chemical adjustment and decontamination andmay include a period of operational (temporary)storage. This initial step provides the best
opportunity to segregate waste streams at source.
This may facilitate recycling within the process
or disposal as non-radioactive waste when thequantities of present radionuclides and theirexemption or clearance levels are known. It also
provides the opportunity to segregate
radioactive waste for different disposal routes.
Prior to treatment of radioactive waste it isessential to characterize the radioactive waste
accurately. Waste characterization is thedetermination of the radiological, chemical andphysical properties of waste to establish theneed for treatment, handling, processing, storage,
or disposal of radioactive materials. Typically,
characterization is helpful in assessing whatmust be done to meet the requirements
regarding transportation and disposal ofradioactive waste. Radiological waste
characterization involves quantifying anddetecting the radiation characteristics of theprincipal radionuclides originated as a result
from the use of radioisotopes or the productionof radionuclides in radiation and nuclear
facilities. The active liquid waste collected fromresearch reactor and radioisotope production
and other research institution in Bangladesh are
generally collect for safe storage at Central
Waste Processing and Storage Facility(CWPSF). The present work was conducted as apart of characterization of radioactive wastes
which are safely stored at waste storage facility.
The liquid waste were analysed for theassessment of the radioactivity concentration.
The Liquid waste which were collected fromdifferent users are now safely stored in theCentral Waste Processing and Storage Facility.
The amount of liquid radioactive waste
generated from different users during 200820015 is shown in Fig1. The liquid wastes aresafely stored at CWPSF in different plastic
containers. After collection of effluents theradionuclide contents analysis andmeasurements of pH and the volume are
performed in the spectrometric laboratory.
The active liquid waste collected from research
reactor and radioisotope production laboratory
was analysed for gamma activity using a HPGe
detector coupled with Genie software. A seriesof activity measurements with a few
representative waste containers were performed.
Before the experiment took place a written
system of work was produced.
70
60
50

Volume (L)Bangladesh

40
30
20
10
0
2008

2009

2010

2011

2012

2013

2014

Year

Fig1. The amount of liquid radioactive waste
generation

2015

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Experimental Methods
To aid risk assessment a non-active run was
initially performed. Contamination was checked
before collecting the sample from waste
container. Latex gloves and leather gloves were
worn during the collection of liquid radioactive
sample. In order to analyse the liquidradioactive sample, one representative sample
from each of the liquid waste container was
taken. Protective clothing, lab coats and masks
were worn before entering the CWPS F in order
to collect the liquid waste sample. The
maximum radiation dose rate and surface
activity of the liquid waste samples wasrecorded using an appropriate survey meter. A
plastic funnel was used to collect the sample
from the active liquid waste container to each of
the sample container. Each sample wascollected in 250 ml plastic pot and plastic tape
was put around the neck of sample pot. Safetymask, polythene, latex and leather gloves wereused to avoid contamination and safety
spectacles were used to protect eyes from liquid
active water drop during the collection ofsamples. A plastic tray was placed at thebeneath of sample container during thecollection of liquid sample in order to avoid thecontamination with liquid radioactive waste in
the floor. The volume of each of the sample pot
was measured before and after the loading of
liquid waste sample. The sample was then put
into a lead pot to take up to the counting room.
The liquid active sample was analysed for γ
activity by using a HPGe detector coupled with
Genie 200. In order to calculate the activity ofthe each liquid waste sample, the energy andefficiency calibration was done using Eu-152
standard source. The standard was placed on the
top of detector within the shielding arrangementin the counting room of HPRWMU. The
spectrometer was set to detect the gamma
energy between 149.77 keV to 3578.36 keV.
The counting time of the each liquid waste
sample was 10,000 sec. The sample was placedin the same position as the standard source for
efficiency calibration. The same geometry waschosen for reference standard source for
efficiency calibration. A plot of energy versus
efficiency is shown in Fig 3. From the sample
net counts activity of the each liquid active
samples and uncertainty of the activity
concentration was calculated using thefollowing standard formula:
𝐶𝐶𝐶𝐶𝐶𝐶

𝐴𝐴 = 𝜀𝜀 ×𝐼𝐼 ×𝑊𝑊𝛾𝛾𝛾𝛾

where, A is the Activity of the sample in Bq l-1,
CPS is the net counts per second, 𝜀𝜀𝛾𝛾 is theefficiency of the γ-ray spectrometer at each
respective γ-ray energy, 𝐼𝐼𝛾𝛾 is the intensity of thecorresponding γ-ray energy, 𝑊𝑊 is the sample
volume.
The uncertainty of the activity:∆𝐼𝐼𝛾𝛾 2
∆𝐴𝐴
∆𝑁𝑁 2
∆𝑤𝑤 2
∆𝑇𝑇 2
�
� �= � � +� � +� � +� �
𝐴𝐴
𝑁𝑁
𝐼𝐼
𝑤𝑤
𝑇𝑇
where, ΔA is the uncertainty of the sample
measurement and ΔN, ΔIγ, Δw and ΔT are theuncertainties of the count rate, gamma-ray
emission probability, sample volume andcounting time, respectively.

Fig 2. Liquid radioactive samples

FNCA
Forum for Nuclear Cooperation in Asia

SampleActivity concentration (Bq l−1) of liquid waste

1.

661.657 keV)
30.15±0.24

60Co (Energy:
1173.228 keV)-

60Co (Energy:
1332.492 keV)
-

5.5

2.

40313.26±0.24

233.12±0.04

235.62±0.06

5.5

3.

85.85 ±0.24

-

-

6.0

4.

8.51±0.34

1335.45±0.03

1314.13±0.06

6.0

5.

22124.93±0.24362.83±0.03

348.89±0.06

6.0Energy (keV)

6.

1.85±0.46

-

-

5.0

1755.42±0.03

1726.99±0.06

5.5

Fig 3. Energy Vs efficiency curve

8.

70.91±0.24

-

-

5.5

9.

7.26±0.53

3943.82±0.03

3930.03±0.06

5.5

10.

8.96±0.33

2341.26±0.03

2311.26±0.06

5.5

113.46±0.04

120.61±0.06

6.5

Efficiency

y = 4.1275x-0.842
R² = 0.9909

137Cs (Energy:

pH

7.

-

11.

-

measured pH of the most of the sample indicatesweak acidic nature of the active liquid wastes.
The dominant gamma lines found in the
spectrum are: 1173.2 keV (0.99), 1332.5 keV(0.99), 661.6 keV (0.85). The identified
radionuclides in the active liquid waste samples
are 60Co and 137Cs respectively.
The specific activity of radionuclides 60Co,
137
Cs in liquid waste samples were analyzed
from dominant gamma energy line by using
Genie 2000 software. Each of the samples wascounted for 10,000 sec and the activities werereported in Bql-1 with uncertainty. The activity
concentration of 137Cs and 60Co in different
liquid waste samples are shown in Fig5. The
calculated activity concentration and theirrespective uncertainties of 60Co & 137Cs are
given in Table1.
45000
40000
35000

Results
The detection and measurement of radionuclide
concentration in the liquid waste samples wereobtained using High Purity Germanium (HPGe)
detector having 20% relative efficiency. The

Activity (Bq/l)

Fig 4. Gamma ray spectra obtained from HPGe
detector with 20% relative efficiency

Concentration of 137Cs

30000
25000
20000
15000
10000
5000
0
-5000
0

1

2

3

4

5

6

Sample Number

7

8

9

10

11

FNCA
Forum for Nuclear Cooperation in Asia

4000

Concentration of 60Co

3500

Activity (Bq/l)

3000

1173.228 keV
1332.492 keV

2500
2000
1500
1000
500
0
2

3

4

5

6

7

8

9

10

11

Sample Number

Fig 5. Activity concentration of 137Cs and60
Co in different liquid waste samples
Table1. Calculated activity concentration of 137Cs and60
Co in different liquid waste samples

The activity concentrations of 137Cs in liquid
waste samples were found to be varied from
(7.26 ± 0.537) Bql-1 to (40313.26 ± 0.24) Bql-1
whilst the activity concentrations of 60Co having
1173.228 keV and 1332.492 keV gamma lines
are ranging from (113.46 ± 0.04) Bql-1 to(3943.82 ± 0.03) Bql-1 and (120.61 ± 0.06) Bql-1
to (3930.03 ± 0.06) Bql-1 respectively. The
information obtained in this work could be used
for the treatment strategy and discharge of thedecontaminated effluents under authorization.

FNCA
Forum for Nuclear Cooperation in Asia

ThailandMs Nanthavan Ya-anantRadioactive Waste Management Center
Thailand Institute of Nuclear
Technology (TINT)

Establishment of New Nuclear Energy
for Peace Act B.E 2559 (2016) in Thailand
Introduction
Thailand has used radioactive materials in the
country since before 1961. In the beginning.
Thailand has no control over this. Thailand
enacted the first Atomic Energy Act for peace in
1961. The utilization of the first nuclear research
reactor in the country started in 1962. Later, theAtomic Energy for Peace Act was amended in
1965. The use of atomic energy purposes was anecessity for the development of the country. It
is important to promote the existence of laws
and regulations on nuclear safety and radiation
safety, as Thailand is a member country of theInternational Atomic Energy Agency. (IAEA).
The aim of this Atomic Energy for Peace Act
1961 was to control the safety aspects of nuclear
research reactor and ionizing radiation.
Since the Atomic Energy for Peace Act 1961 is
not present so it is need to be updated.
In accordance with the international standards
for the control and prevention of radiation
hazards, so the Nuclear Energy Act B.E.2559
was enacted in 2016. In line with modern
nuclear technology, which has developed andutilized in the country and to comply withinternational rules and regulations, including
nuclear and radiation treaties, such as nuclear
oversight, covering safety, security andsafeguard as well as nuclear facilities. The new
Act has been in force since February 2017, inorder to regulate radioactive material, nuclear
material, radiological generator, including

radioactive waste. The new Act will promote thesafety, security and safeguard in nuclear
application in Thailand.
Main important points of Nuclear Energy for
Peace Act B.E.2559
In order to use of radioactive material, nuclear
material and radiation safely, importers,
exporters, manufacturers, users and otherstakeholders need to know and comply with therules and regulations. In this New Nuclear
Energy for Peace Act B.E.2559, the main issuesare:• Regulation on Radioactive Material• Regulation on Radiological Generators• Regulation on Nuclear Material• Regulation on Nuclear Facilities
Nuclear Energy for Peace Act B.E.2559 has
been publishing in the Royal Gazette Vol. 133,
No. 67 A, dated 5 August 2016, as shown in
Figure 1. The Nuclear Energy for Peace Act
B.E.2559 has been in force since February 1,
2017. The Ministerial Regulations , Nuclear
Energy Commission Orders and Office ofAtoms for Peace Rules under this Act shall be inplace after 270 days after the enforcement of
this Nuclear Energy for Peace Act B.E.2559.

Fig.1 Nuclear Energy for Peace Act B.E. 2559 [1]FNCA
Forum for Nuclear Cooperation in Asia

Summary of Nuclear Energy for Peace Act
B.E.2559
There are 14 Chapters 152 Articles in the new
Nuclear Energy for Peace B.E.2559 as
followings,
Chapter 0. Preamble and Definitions (Art. 1-5)

Chapter 1. General Provision for Nuclear
Energy (6-8)
Chapter 2. Nuclear Energy for Peace
Commission (9-17)
Chapter 3. Radioactive Materials andRadiation Generating Machine (18-35)
Chapter 4. Nuclear Materials (36-44)
Chapter 5. Nuclear Facilities (45-64)
Chapter 6. Radioactive Waste (75-83)
Chapter 7. Spent Fuel (84-87)
Chapter 8. Safety, Security and Safeguard(88-97)
Chapter 9. Transportation (98-99)
Chapter 10. Radiation and Nuclear
Emergency (100-102)
Chapter 11. License Revocation (103-104)
Chapter 12. Appeal (105-106)
Chapter 13. Officer (107-114)
Chapter 14. Punishment (115-144)
Chapter Transitory Provision (145-152)
Radioactive waste management and spent
nuclear fuel management are in the chaper 6 and
7 respectively.
Chapter 6. Radioactive Waste Management(RWM)• Import of Radioactive Waste (RW) is
prohibited, except re-import of reprocessingof Spent Fuel and conditioning of RW.
• Export/Import of RW should be authorized
by OAP Secretary General. Requirement
and condition specified in Ministerial
Regulations.

• Discharge the RW into the environment isprohibited, except the RW which radio
activities and half-life are specified in
Ministerial Regulation.
• RW producer must be responsible themanagement of RW. Requirement andcondition specified in Ministerial
Regulations.
• RWM service provider shall be got thelicenses for siting, construction,commissioning of RWM Facilities andActivities.
• DRSR which more than 5 year no-longer
use shall be treated in the manner ofRadioactive Waste
Chapter. 7 Spent Nuclear Fuel (SNF)• Export and Import of SNF are prohibited,
except re-import of reprocessing of SNF.
• Export/Import of RW should be authorized
by SG. Requirement and condition
specified in Ministerial Regulation.
• SNF shall be storaged as specified in Final
Safety Analysis Report. Requirement andcondition specified in Ministerial
Regulation.
• Nuclear facilities have to store their own
SNF safely and securely as described intheir safety analysis reports (SARs) unless
SF is• transferred to government entity
responsible for SF storage, or• exported to be reprocessed, or• exported (sent back) to manufacturers
or lessors
Reference[1] Nuclear Energy for Peace Act B.E. 2559, theRoyal Gazette Vol. 133, No. 67 A, dated 5
August 2016,

The FNCA RS&RWM Project Leaders
R.O. Korea

Australia

Dr Hyuncheol Kim
Senior Researcher
Environmental Radioactivity
Assessment Team
Korea Atomic Energy ResearchInstitute (KAERI))

Ms Lynn Tan
Project Transition Leader
Waste Projects and Strategic Planning
Australian Nuclear Science andTechnology Organisation (ANSTO)

Banglades

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

China

Malaysia
Dr Norasalwa Zakaria
Manager, Waste and Environment
Technology Development Centre,
Waste and Environmental Division,
Malaysian Nuclear Agency (Nuclear
Malaysia)

Mongolia
Dr Zhang Jintao
Deputy Director General,
Department of Safety, Protection andQuality,China National Nuclear
Corporation (CNNC)

Indonesia

Ms Oyuntulkhuur Navaangalsan
Head,
Radiation Safety Council
Nuclear Energy Commission

Philippines
Mr Husen Zamroni
Director of Center forRadioactive Waste Technology, National
Nuclear Energy Agency of Indonesia(BATAN)Japan

Ms Maria Visitacion B. Palattao
Head, Regulations and Standards
Development Section, Philippine
Nuclear Research Institute (PNRI)

ThailandMs Nanthavan Ya-anantHead, Radioactive Waste Management
Section,Radioactive Waste Management
Center,Thailand Institute of Nuclear

Prof Toshiso Kosako
Professor Emeritus,
The University of Tokyo

Kazakhstan

Vietnam
Mr Yevgeniy Tur
Technical and Economic Assessment
Department
National Nuclear Center (NNC) of the
Republic of Kazakhstan

Dr Pham Quang Minh
Director, Department for Planning and
R&D Management,
Vietnam Atomic Energy Institute(VINATOM)
Technology (TINT)

FNCA
Forum for Nuclear Cooperation in Asia

FNCA Secretariat
International Affairs & Research Department,
Nuclear Safety Research Association (NSRA)
5-18-7 Shimbashi, Minato-ku, Tokyo, 105-0004,
Japan
TEL: +81-3-5470-1983
FAX: +81-3-5470-1991
E-mail: fnca@fnca.mext.go.jp