Kenaf (Hibiscus cannabinus L.) is one of the world’s most economically important fiber crops particularly in Asia-Pacific region. Mutation induction is a method to increase genetic divergence associated with selection, recombination, or a combination of these approaches in plant breeding. One of preliminary procedures for an execellent mutation breeding program is the radiosensitivity study to determine the optimal doses for irradiation. A total of 10 different doses of acute gamma rays (0, 100, 200, 400, 800, 1000, 1200, 1500, 1700 and 2000 Gy) from ceasium-137 source were applied to the seeds of V 36 kenaf variety. The irradiated seeds including the control were planted in trough for 30 days. The gamma irradiation effects on several parameters such as seedling survival percentages, plant height, root length, shoot fresh weight, root frest weight, shoot dry weight and root dry weight were analysed. From the radiosensitivity curve, the LD50 and LD25 values were estimated at 810 and 310 Gy, respectively. Two doses (200 Gy and 300 Gy) based on LD25 were chosen for evaluating the effects of gamma irradiation on morphological traits in M1 generation. Irradiated and non-irradiated seeds were planted in the field at Beseri, Perlis. About 10 phenotypic traits of irradiated plants were observed and evaluated against the controls. Cluster analysis on M1 progenies showed that the mutation could be classified into eight genotypic groups. The first two components from principal component analysis explained about 77.99% of variation. Number of seeds per pod, weight of seeds per plant and dry stem biomass play an important role in explaining the variation since they showed positive correlated values for the first component analysis. Morphological changes such as flower shape, flower colour, and leaf shape were also observed in M1 generation. The findings of this study are important in determining the effectiveness of these doses in generating mutations on kenaf plant and subsequent breeding program to develop new kenaf varieties with enhanced quality traits.
Nuclear grade (high-purity) graphite for fuel element and moderator material in Advanced Gas
Cooling Reactors (AGR) displays large scatter in strength and a non-linear stress-strain response from the damage accumulation. These responses can be characterized as quasi-brittle behaviour. Current assessments of fracture in core graphite components are based on the linear elastic approximation and thus represent a major assumption. The quasi-brittle behaviour gives challenge to assess the real nuclear graphite component. The selected test method would help to bridge the gap between microscale to macro-scale in real reactor component. The small scale tests presented here can contribute some statistical data to manifests the failure in real component. The evaluation and choice of different solution design of biaxial test will be discussed in this paper. The ball on-three ball test method was used for assessment test follows by numerous of analytical method. The results shown that biaxial strength of the EY9 grade graphite depends on the method used for evaluation. Some of the analytical methods use to calculate biaxial strength were found not to be valid and therefore should not be used to assess the mechanical properties of nuclear graphite.
The remote measurements of radiation level at an identified location, are not only important for
collecting data or monitoring radiation level per se, but also crucial for workers who deal with
radiation sources. A device for checking an on-site radiation level has been developed quite a
long time ago under the name of Geiger Muller and widely known as a Geiger counter. The
reading of the output can be seen on the device on-site and on real-time basis. Nowadays, with
the fast evolution of computer and networking technology, those reading not only can be read
real-time but also from a remote location that makes workers able to enter the risky area more
safely. The collected data reading also can be analyzed for predicting the future trending
pattern. The data is transferred from the monitoring devices to a server through a network. This
paper discusses about several critical issues on the design, implementation and deployment that
relates to the devices, interface programs, hardware and software that allow all parameters such
as radiation levels reading and the timestamp of the data-logging can be collected and stored in
a central storage for further processes. The compatibility issue with regards to technology
change from the previous system will also be discussed. The system has many advantages
compared to previous system and conventional method of doing the area monitoring in term of
sustainability and availability.
This study evaluates the ImageJ software as dosimetric tools for analyzing the film dosimeter in high energy photons and electrons. The percentage depth dose of photons of 6 and 10 MV and electrons of 6 and 9 MeV were measured using the Gafchromic EBT2 film dosimeter. The films were scanned and analyzed using the Verisoft software and ImageJ. The beam profiles at nominal photon and electron beam parameters were also evaluated using the two methods. The PDD measured in ImageJ at high energy photons were in good agreement within 1% percentage of discrepancy at all depths in comparison to the Verisoft software. The PDD measured in ImageJ at high energy electrons also showed good agreement to Verisoft software within 8% percentage of discrepancy at all depths. The measured flatness of beam profiles at Dmax, R50, R80 and R90 in ImageJ were also in good agreement to Verisoft software with flatness value between 4 and 8%. The results indicated the suitability of ImageJ software as dosimetric tool for analyzing EBT2 film dosimeter at high energy photon and electrons.
The mass attenuation coefficients of solid water phantoms, Perspex® phantoms and Rhizophora spp. particleboards were determined by using Compton scattering technique measured using Ludlum configuration. The gamma energy of 137Cs sealed source were measured at 30°, 45°, 60° and 75° angles providing scattered gamma energies between 337.72 and 564.09 keV. The mass attenuation coefficients of solid water and fabricated Rhizophora spp. particleboards were the nearest to XCOM values of water with average percentage of discrepancies of 6.8% and 5.9%, respectively. The results indicated similar attenuation properties of solid water and fabricated Rhizophora spp. particleboards and the suitability of the Ludlum configuration to determine the mass attenuation coefficient of materials using Compton scatter technique.
An appraisal of the regional reconnaissance geophysical, geochemical and geological data obtained under the Central Belt Project in 1977 – 1978 appeared to constitute favourable uranium exploration targets. Follow-up surveys conducted until the year 1990 have proposed the exploration area to be divided into three transects. Transect 1 covers the western part of the state of Kelantan, northwest Pahang and the eastern half of Perak. Transect 2 covers southeastern tip of Perak, west Pahang, eastern half of Wilayah Persekutuan, eastern portion of Selangor, and the northwest portion of Negeri Sembilan. Transect 3 covers central Kelantan, northwestern of Terengganu and northern portion of Pahang. Results of the study indicate that the Main Range, Bujang Melaka, and Bukit Tinggi Plutons are most fertile with uranium spectrometric rock values range from 13 to 25 ppm. Further investigations to zero down the target areas for uranium mineralization are strongly recommended over these areas.
Semiconductor thin films Copper Tin Selenide, Cu2SnSe3, a potential compound for solar cell applications or semiconductor radiation detector were prepared by thermal evaporation method onto well-cleaned glass substrates. The as-deposited films were annealed in flowing purified nitrogen N2, for 2 hours in a temperature range from 100˚C to 500˚C. The structure of as-deposited and annealed films has been studied by X-ray diffraction technique. The semi-quantitative analysis indicated from Reitveld refinement show that the samples composed of Cu2SnSe3 and SnSe. These studies revealed that the films were structured in mixed phase between cubic space group F-43m (no. 216) and orthorhombic space group P n m a (no. 62). The crystallite size and lattice strain were determined from Scherrer calculation method. The results show that increasing in annealing temperature resulted in direct increase in crystallite size and decrease in lattice strain.
The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial
criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of
basic nuclear research, manpower training, and production of radioisotopes. This
paperdescribes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP);
focusing on the application of the developed reactor 3D model for criticality calculation,
analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D
continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full
model of the TRIGA reactor. The consistency and accuracy of the developed RTP MCNP model
was established by comparing calculations to the experimental results and TRIGLAV
code.MCNP and TRIGLAV criticality prediction of the critical core loading are in a very good
agreement with the experimental results.Power peaking factor calculated with TRIGLAV are
systematically higher than the MCNP but the trends are the same.Depletion calculation by both
codes show differences especially at high burnup.The results are conservative and can be
applied to show the reliability of MCNP code and the model both for design and verification of
the reactor core, and future calculation of its neutronic parameters.
The Malaysian 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the calculation of neutron flux and power distribution in PUSPATI TRIGA REACTOR (RTP) 14th core configuration. The 3-D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA core and fuels. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data as well as S (α, β) thermal neutron scattering functions distributed with the MCNP code were used. Results of calculations are analyzed and discussed.
The Malaysian’s PUSPATI TRIGA Reactor (RTP) achieved its initial criticality on June 28, 1982. The reactor is designed to effectively implement various fields of basic nuclear research, manpower training, and production of radioisotopes. Several past activities on neutronics modelling development and validation of the RTP were carried out using Monte Carlo Code MCNP. In this work, the developed model was used to characterise in-core and beam-ports irradiation facilities of the reactor. The thermal and fast neutron flux distributions in these facilities were determined using MCNP mesh tally method. It was found that the flux as well as its spectral characteristics depended very much on the position of the irradiation facility in the reactor core or in the beam-ports. The maximum neutron flux was found to be in the Central Thimble facility with 1.98E13 nv of thermal neutron. The thermal-to-total flux ratio varies significantly from 0.41 for the in-core facility, 0.58 in the reflector and up to 0.88 in the beam-ports.
IAEA Soil-6 is a reference material with a certified value for 226Ra fall between 69.6 – 93.4 Bq/kg at 95% confidence level. This material has been used as a sample and performed repeat measurement weekly between years 2006 – 2009 using a same gamma spectrometry system. The activity concentration of this material is calculated automatically using the operational commercial software and compared with activity obtained from the manual calculation. Study found that only 76.9%, 64.1%, 56.3%, and 79.3% of the results from the software calculation lie within the confidence level for year 2006, 2007, 2008 and 2009, respectively. However, u-score calculation revealed that 94.9 %, 89.7%, 79.2% and 84.9% data set have no significant bias (u < 2.58) for year 2006, 2007, 2008 and 2009, respectively. On the other hand, all manual calculation data were found to be within the 95 % confidence level. Factors suspected to cause differences between these two approaches were discussed here. Manually peak search, marking and calculation still remains as the preferred option for calculating the gamma radionuclides activity unless limitations of the spectrum analysis software, as described in this paper can be resolved/improved upon.
Natural radioactivityin thesurface soil from thevicinity of Lynas Rare-earth plantin Kuantan, Pahanghad been studied with aim of evaluating environmental radioacitivityof thatarea. The radioactivities of the samples have been measured usinga low-background HPGe gamma detector. The concentration of 226Ra ranged found to be between 8.7 –76.9 Bq/kg (mean 33.7 Bq/kg); activity of 238U from 8.7 to 80.5 Bq/kg (mean 36.9 Bq/kg); as for 232Th, it ranged from 6.2 to 121.5 Bq/kg(mean 58.2 Bq/kg)and that of 40K,the values was from 19.1 to 778.9 Bq/kg (mean of 221.3 Bq/kg). Activity standard deviation for226Ra, 238U, 232Th and 40K were 14.2 Bq/kg, 14.7 Bq/kg, 26.1 Bq/kg and 185.6 Bq/kg, respectively. To evaluate the risk of radiation hazard to human, several parameters, i.e. the radium equivalent activity, External hazard index, Representative level index and dose rate in air from the terrestrial natural gamma radiation were calculatedwith the mean valuesfound to be136.7 Bq/kg, 0.4, 1.0, 65 nGy/h, respectively. The data were discussed and compared with those obtained from the previous studies given in the literatures. Thisstudy found that themean dose equivalent received by an individual at the vicinity area (outdoor) of Lynas plant was estimated to be 82.0 Sv/yr, whichis far below the annual dose limitof1,000 Sv/yrfor general public as stipulated in the national act.
Assessment of source and sediment quality was carried out on marine sediments collected from the Tuanku Abdul Rahman National Park. Enrichment factors (EF), pollution load index (PLI) and geo-accumulation index (Igeo) were used to identify the sources of pollution, degree of contamination and sediment quality, respectively. Elemental analyses of marine sediment samples were performed by using the Instrumental Neutron Activation Analysis (INAA). Results from the Tunku Abdul Rahman National Park of Sabah indicated that most of the elements are considered to be from lithological or natural origin with EF values of less than 2 except for As (10 stations), Cr (3 stations), Lu (5 stations), Mg (2 stations), Sb (6 stations) and U (3 stations). For the sediment quality, most of the study area can be categorised as unpolluted for most of the elements (Igeo value < 2) except for As, Cr, Lu, Mg, Sb and U. A few study areas were slightly low contaminated with As, Cr, Lu, Mg, Sb and U. The contamination of As, Cr, Lu, Mg, Sb and U in the study area can be categorised as moderate with Igeo values ranged from 1 to 2. Meanwhile, the results of PLI value for sediment were ranged from 0.93 to 1.47 (PLI < 50) indicating there are not required to perform drastic rectification measures for the screening of the elements in the Tunku Abdul Rahman Park. Overall, assessment of the sediment quality at the Tunku Abdul Rahman National Park showed a few elements such as As, Cr, Lu, Mg, Sb and U were slightly enriched while most of the elements were similar to background values.
In industrial plants such as electricity generating, petroleum, chemical and petrochemical plants, pipelines are used extensively to transport liquid from one location to another. In radiation vulcanization of natural rubber latex (RVNRL) plants, pipelines are also used to transport latex to storage tank. During one of its maintenance activities, a pipeline intelligent gauge (PIG) that was used to monitor pipe integrity jammed inside the pipe causing interruption to its operation or loading activities. Sealed source technology was utilized to determine the location of jammed PIG in the pipeline. Fast neutrons from a 50 mCi Americium Beryllium (AmBe241), with energy range between 0.5 to11 MeV, were used for the study. Helium 3 (He3) detector was used to detect slow neutrons having a range of energy of 30 eV- 0.5 MeV. The investigation was carried out using neutron backscatter technique scanner. By adopting back-scattered technique, the location of jammed PIG in the pipeline has been successfully determined.
The total mass attenuation coefficients (μ/ρ) of stainless steel (SS316L) and carbon steel (A516) that are widely used as petrochemical plant components, such as distillation column, heat exchanger, boiler and storage tank were measured at 662, 1073 and 1332 keV of photon energies. Measurements of radiation intensity for various thicknesses of steel were made by using transmission method. The γ-ray intensity were counted by using a Gamma spectrometer that contains a Hyper-pure Germanium (HPGe) detector connected with Multi Channel Analyzer (MCA). The effective numbers of atomic (Zeff) and electron (Neff) obtained experimentally were compared by those obtained through theoretical calculation. Both experimental and calculated values of Zeff and Neff were in good agreement.
The glow curve in TLD-100 was compared by applying long preheat time, short preheat time
techniques and without preheat technique before the TLD readout. Fading effect of the TLD signal
upon certain storage time with long preheat time (100°C, 10 minutes using the oven) and short
preheat time techniques (100°C, 10 seconds using the reader) were also studied. 15 TLD-100 chips
were used with 3 of the TLD chips were used for measuring background radiation. 12 TLD chips
were annealed, irradiated, preheated long and short preheat time techniques) and analyzed. The TL
signals output from TLD chips of without preheated were used as the control. Two sets of data were
taken using TLD chips irradiated with 6 MV and 10 MV photon beams. TL signal output was
recorded the highest for short preheat time, followed by long preheat time and no preheating. The
TL signal loss upon certain storage time was also reduced when short preheat time technique was
applied. By applying long preheat time technique the low temperature peak in the glow curve was
completely removed for both energies. Whereas, TLD chips exposed to 6 MV and with short preheat
time technique the low temperature peak did not disappear completely but decreased in intensity as
compared to the control data by 19.80%, 37.69%, 48.19% and 100% at 24, 48, 72 and 96 hours
after exposure prior to readout, respectively. Meanwhile, for 10 MV photon beam with short
preheat time, the small peak intensity was reduced by 19.58% for readout at 24 hours after
irradiation and 100% for 48,72 and 96 hours delayed time prior to readout. It was observed that
the TLD-100 was highly dependent on preheat heating time before readout. Short preheat time
technique was able to reduce post irradiation fading of TLD-100 dosimeters
Pulse Shaping Amplifier (PSA) is an essential component in nuclear spectroscopy system. This
amplifier has two functions; to shape the output pulse and performs noise filtering. In this paper,
we describe the procedure for the design and development of a pulse shaping amplifier which can
be used in a nuclear spectroscopy system. This prototype was developed using high performance
electronics devices and assembled on a FR4 type printed circuit board. Performance of this
prototype was tested by comparing it with an equivalent commercial spectroscopy amplifier (Model
Silena 7611). The test results showed that the performance of this prototype was comparable
to the commercial spectroscopic amplifier.
Radiation Monitoring System aims to limit the exposure dose to personnel to the lowest level
referring to the concept of ALARA (As Low As Reasonably Achievable). Atomic Energy Licensing
(Basic Safety Radiation Protection) Regulations 2010 is useful to control employee and public
radiation protection program and guideline. This paper discussed the methodology and
implementation of the radiation protection program at PUSPATI TRIGA Reactor (RTP) which is
implemented in Nuklear Malaysia, Complex Bangi.
The combination of irradiation and biological technique was chosen to study COD, BOD5 and colour removal from textiles effluent in the presence of food industry wastewater. Two biological treatments, the first consisting a mix of non irradiated textile and food industry wastewater and the second a mix of irradiated textiles wastewater and food industry wastewater were operated in parallel. Reduction percentage of COD in textiles wastewater increased from 29.4% after radiation to 62.4% after further undergoing biological treatment. After irradiation, the BOD5 of textiles wastewater was reduced by 22.1%, but reverted to the original value of 36mg/1 after undergoing biological treatment. Colour had decreased from 899.5 ADMI to 379.3 ADM1 after irradiation and continued to decrease to 109.3 ADMI after passing through biological treatment.
Sludge and scales produced during oil and gas production contain enhanced naturally occurring
radioactive material (NORM). Sludge and scales are under the jurisdiction of Department of
Environment (DOE) and also Atomic Energy Licensing Board (AELB). AELB has issued a
guideline regarding the disposal of sludge and scales as in its guideline (LEM/TEK/30 SEM.2,
1996). In this guideline, Radiological Impact Assessment (RIA) should be carried out on all
proposed disposals and has to demonstrate that no member of public will be exposed to more than
1 mSv/y. This paper presented RIA analysis using RESRAD computer code for the disposal of
treated sludge. RESRAD (RESidual RADioactive) developed by Argonne National Laboratory is to
estimate radiation doses and risks from residual radioactive materials. The dose received by the
member of public is found to be well below the stipulated limit.