International Journal of Nuclear Energy Science and Technology (13 papers in press)
Comparison of TRIGA reactor steady-state thermal-hydraulic predictions by COMSOL multiphysics with experimental data
by Ahmed K. Alkaabi, Jeffrey King
Abstract: This paper presents United States Geological Survey TRIGA one- and multiple-channel thermal hydraulic (TH) models developed using the COMSOL code to examine the effects of coolant cross-flow on coolant, cladding, and fuel temperatures. There are considerable variations in the profiles of the coolant axial temperatures and outlet temperatures as predicted by multiple-channel model from those predicted by the one-channel model. The one-channel model forecasts that the temperature of the coolant within the fuel rings increases axially with the height of the core, whereas the temperature of the coolant predicted by the multiple-channel model increases as a function of core height in the B-, C-, and D-rings, peaks and then reduces within the E-, F-, and G-rings. Within the multiple-channel model, the coolant appears to flow from the cores outermost opening at the lower side to the centre of the core. Finally, predictions of all models are benchmarked with the experimental data.
Keywords: TRIGA reactors; multiple-channel models; thermal hydraulic analysis.
CFD analysis of primary and secondary sodium flows and associated heat transfer on performance of an intermediate heat exchanger in LMFBR
by SUYAMBAZHAHAN Sivalingam, Sundararajan Thirumalai, Sarit Kumar Das
Abstract: The occurrence of non-uniform temperature in secondary sodium flow in an Intermediate Heat Exchanger (IHX) causes uneven expansion in the long stainless-steel tubes and differential thermal stresses in components. CFD analysis was carried out to find the pressure drop and heat transfer between primary and secondary sodium flow in IHX of a liquid metal-cooled fast breeder reactor (LMFBR) using ANSYS 18.1 software. A similar anisotropic porous technique was used for both primary and secondary sodium flow inside and outside the bank of tubes. A decoupled analysis was performed with prescribed heat source/sink distribution. The distributors and baffles are attached to achieve uniform flow and heat removal in secondary sodium flow. A coupled analysis of the primary and secondary sodium heat exchange was also performed by using the flow field obtained with the distributor. The flow distributor reduced the radial temperature variation < 20 K within the bundle region. A 3D sectoral model was analysed and estimated the effects of a mixing device in the secondary flow. The temperature difference again reduced to 7 K at the secondary flow exit, by mixing hot sodium of outer region with that of the inner region. Thus, achieved nearly uniform temperature in the secondary flow of the IHX in LMFBR.
Keywords: IHX performance; LMFBR; flow mal-distribution; decoupled; coupled and combined CFD analysis.
The power loss requirement in the non-thermal 3He-3He fusion plasma
by Javad Bahmani
Abstract: The control of power loss is the most important issue in the plasma and it should be minimised by suitable selection of design parameters. This research is done with the aim of determining power loss and the parameters affecting the 3He-3He plasma with isotropic non-Maxwellian electron velocity distribution function (EVDF). In order to investigate the power loss requirement, bi-Maxwellian EVDF is considered. The results show that the rate of entropy production decreases as the electron density decreases and collision time increases. The ratio of the minimum power loss to the fusion power decreases as the electron temperature, electron, and helium-3 densities decrease, and the ion temperature increases. It is favorable to decrease the electron temperature to below the thermal limit. Moreover, the reduction of the electron temperature increases the fusion power.
Keywords: non-thermal plasma; entropy; power loss.
Validation study of reactor physics code WIMSD-5B based on evaluated nuclear data libraries for VVER core calculations by benchmarking VVER critical experiments of light water reactors
by K. M. Zaheen Nasir, Tanaya Chakma, Benozir Ahmed, Mohammad Jahirul Haque Khan
Abstract: This work focuses on the study of integral parameters of VVER-type critical experiment lattices of LWRs for VVER core calculations with the purpose of validation of the lattice transport code WIMSD-5B based on evaluated nuclear data files ENDF/B-VII and JENDL-3. These lattices are the standard benchmarks for testing reactor physics methods (codes) and libraries. The integral parameters of these lattices were calculated using WIMSD-5B based on the aforesaid libraries and were compared with the measured values as well as evaluated calculations of different libraries. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment values and evaluated calculations. This analysis reflects the validation of WIMSD-5B and associated libraries. Furthermore, the group constants study also enriches this validation work. Therefore, this study is essential to implement further neutronics analysis and core calculations of the VVER reactors using same code and libraries.
Keywords: validation; integral parameters; benchmarking; WIMSD-5B code; VVER; criticality lattice.
Design, analysis and experimental qualification of gripper bellows for FBTR control rod drive mechanism
by S. C. S. P. Kumar Krovvidi, Ramesh E, Iyyappan R, Kolanjiappan A, Chandramouli S, Sreedhar B K, Sureshkumar K V, Raghupathy S, Bhaduri A K
Abstract: The design of welded-disc bellows is not addressed in any design codes or standards. This paper discusses the design, analysis and experimental validation of the gripper bellows used in control rod drive mechanisms of fast breeder test reactor. The material of construction of the bellows is 0.125 mm thick AM350-SCT1000. The design fatigue curve of the material at 803K is not available in open literature. The fatigue curve of the material (Coffin-Mansons relation) was generated from the tensile testing data using four-point correlation method and universal slopes method. Elasto-plastic finite element analysis of the bellows was carried out and fatigue life of the bellows was estimated numerically. A factor of safety of 20 on the number of cycles was used as suggested in codes such as ASME section-III & RCC-MR. Experimental qualification of the bellows as per ASME section-III, appendix-II was carried by testing of the bellows in sodium at 803 K.
Keywords: welded disc bellows; AM350; in-sodium testing; Coffin-Manson plot; four-point correlation; finite element analysis.
Spallation reaction study to improve cross-section measurements of fission products in nuclear waste using Cs-137 on proton or deuteron ranging from 0.1 to 2.4 GeV
by Abdessamad Didi, Hamid Amsil, Hamid Bounouira, Khalid Laraki, Hamid Marah, Hassane Dekhissi, Mohammed Yjjou
Abstract: The aim of this research was to study the spallation reaction of a caesium-137 target using a beam of protons or deuterons for the transmutation of nuclear waste, as well as evaluate the differences in the production cross section of the secondary spallation products, such as neutrons, protons, deuterons, pions (pi_+, pi_-, pi_0), helions, tritons, and alphas. The work presented in this article provides the necessary scientific evidence for confidently implementing the MCNP-calculated transmutation of caesium-137 using a spallation reaction. In our research we evaluated and improvised the different physical characteristics of the fission products of the caesium-137 target during a spallation reaction.
Keywords: transmutation; waste energy; spallation; caesium-137; MCNP; cross section; Monte Carlo.
Real time sub-assembly identification through IMU data fusion with vision sensor for an inspection system
by Thirumalaesh Ashokkumar, N.A. Nibarkavi, S. Joesph Winston, Joel Jose, Rathika P D
Abstract: Prototype Fast Breeder Reactor (PFBR), is a two loop, sodium cooled, pool type reactor. The PFBR reactor core is made up of sub- assemblies holding core material in hexagonal lattice. The inspection of core internals is essential to ascertain the structural integrity of the core components. During the commissioning of the reactor, this inspection will be of importance for the qualification of the core components after the assembly and integration, which will give confidence in successfully running the reactor. The Reactor Core Viewing System in Room Temperature (RCVS-RT) is a system that has been developed to aid in this process for the inspection of the reactor core components. RCVS-RT is designed to introduce a vision probe through the reactor top shielding through the observation port into the core top and then into the extracted sub- assembly slot to reach up to the grid plate top for inspection. The RCVS has a linear radial, rotational theta and an azimuth linear vertical axis to articulate the camera pipe during deployment. The RCVS-RT is tested and qualified on a mock test setup to qualify the system on a 19 dummy sub-assembly arrangement. Since the orientation is completely lost during the deployment of camera probe, it is imperative to implement a feature to overlay the orientation information into the vision data for assisting the user during inspection. This work explores fusing non- contact sensors to a vision sensor to achieve orientation recognition of the RCVS- RT. The use of non-contact sensors helps in maintaining sterile core components without having the inspection tools interacting with them. It also ensures data that is reliable. A 10 DOF GY-87 IMU sensor that comprises of a HMC5883L is used with the vision sensor. The sensor module has a Magnetometer, MPU6050 Accelerometer and Gyroscope and a BMP180 Air pressure sensor. Using the orientation of the RCVS-RT and the numbering of the extracted sub assembly, the numbering of the adjacent sub-assemblies is deduced from the core geometry. This allows for a straight forward identification of the subassemblies in the core. Another simpler test setup was designed to check the sensor data fusion to have orientation overlaid on vision data.
Keywords: PFBR; RCVSRT; data fusion; orientation identification; sub-assembly numbering algorithm.
Assessment of annual effective dose and excess lifetime cancer risk due to alpha emitters in some grain samples in Kerbala Governorate, Iraq
by Zahraa Saad Hamzah, Abdalsattar Kareem Hashim, Ali Abojassim
Abstract: This study aims to measure radon concentration in some types of grain samples of Kerbala Governorate. The measurements of radon concentration were carried out using a CR-39 detector. Also, radium-226 and uranium-238 were calculated to depend on radon concentrations in the samples of the present study. Furthermore, annual effective dose (AED) and excess lifetime cancer risk (ELCR) due to alpha emitters (222Rn, 226Ra, and 238U) in all samples were determined. The results show that radon concentrations in air space varied from 0.12 to 40.56 Bq/m3, while the results of radium and uranium varied from 0.14 to 46.25 mBq/kg and from 0.002 to 0.706 Bq/kg, respectively. Also, the average value of the total of AED and ELCR was 0.402
Keywords: alpha emitters; radon concentrations; cancer risk; and kerbala governorate.
Determination of radiological hazards due to alpha emitters from ceramics used in Iraq
by Sara SalihNayif, Elham Jasim Mohammed, Abdalsattar Kareem Hashim, Ali Abojassim, Hussien Abid Ali Bakir Mraity
Abstract: The sealed can technique was used in this work to determine the amount of radioactivity (alpha emission) of imported ceramic tiles that are used in different kinds of building in Iraq. The resulting data showed that the radon concentration varied from 22.105 to 302.482 Bq/m3 with an average of 162.293 Bq/m3. The effective radium content ranged from 0.079 to 1.087 Bq/kg with an average value of 0.583 Bq/kg. The uranium concentration varied from 1.192 to 16.313 Bq/kg with an average value of 16.313 Bq/kg. After obtaining those results and comparing them with the global average and permissible limits recommended by international scientific agencies, such as ICRP and UNSCEAR, it was found that the considered ceramic samples are safe for local use.
Keywords: alpha emitters; ceramic; radiological hazards; closed-can technique.
Xe and Kr extraction for Th-U sustainable ICMSR fuel
by Iza Shafera Hardiyanti, A. Suparmi, Andang Widi Harto
Abstract: the Innovative Compact Molten Salt Reactor (ICMSR) is a nuclear reactor designed to use thorium as the main fissile fuel to achieve sustainable fuel resources. ICMSR has the inherent safety required as an advanced reactor. This reactor uses liquid fuel salt. The fuel contains NaF-ThF4-UF4 (75-19.4-5.6) % mole of fuel salt with 19.75% uranium-235 enrichment. ICMSR uses graphite as moderator, Hastelloy-N as reactor vessel and NaF-KF (50-50) % mole as intermediate coolant. Fission yields produced by ICMSR include Xe and Kr. These isotopes need to be removed from the reactor because they reduces fuel utilisation. This paper describes the effect of the extraction of Xe and Kr isotopes on the criticality of ICMSR. Calculation is done by using MCNP6. The results show that the extraction of Xe and Kr on ICMSR increased the criticality and produced 2.661E+08 Ci actinide fission products
Keywords: ICMSR fuel; thorium sustainable fuel; Xe and Kr extraction; criticality; fission yield.
Background radiation in primary schools of Al-Najaf City, Iraq
by Rukia Jabar Dosh, Ali K. Hasan, Ali Abojassim
Abstract: In this work, background radiation (absorbed dose rate) was measured in the air of buildings in primary schools in Al-Najaf governorate, Iraq, using a portable radiation dosimeter (Inspector Exp. USA). The annual average dose (AED) and excess lifetime cancer risk (ELCR) were calculated in all schools under study. Background radiation results were drawn using a geographic information system (GIS) technique. The results show that the average values of dose rate (
Keywords: background radiation; dose rate; cancer risk; primary schools; GIS technique; Najaf city.
Radiogenic heat production from natural radionuclides in sediments of the Tigris River in Mosul City, Iraq
by Laith Ahmed Najam, Sheamaa T. Al-Dbag, Taha Y. Wais, Howaida Mansour
Abstract: In this paper, the concentration of the activity of natural radionuclides is represented by 238U, 232Th, and 40K. This concentration was measured by gamma spectroscopy, which was used to determine and calculate radioactive heat production for all radionuclides and their distribution pattern in sediment samples from the Tigris River in Iraq. The results showed that the activity concentration of these radionuclides was within the allowed limits by the UNSCEAR. Field observations show that the sediments occurred due to the weathering of pre-existing sedimentary rocks that make up the area's geology. The estimated average radiogenic heat production rate was 0.3634
Keywords: heat production; Tigris River; sediments; radionuclides; gamma spectrometry.
BAEC TRIGA research reactor: 35 years experience
by Md. Abdul Malek Soner, Abdullah Al Mahmud, Md. Moazzem Hossain, Md. Sayed Hossain, Md. Mobasher Ahmed, Md. Rakibul Hasan, Md. Bodhroddoza Shohag, Nusrat Jahan, Mohammad Mezbah Uddin, Ashraful Haque
Abstract: BAEC TRIGA Research Reactor (BTRR) is the only nuclear reactor in Bangladesh. Its a valuable tool for a wide variety of research accomplishments and serves as an excellent source of neutrons. BTRR has been operating since 14 September 1986, and is used in various fields of research and utilisation. BTRR is continuously playing its part in the development of the nuclear sector in Bangladesh. Excluding some incidents, BTRR operation has been successfully carried out till now. The few incidents, such as a decay tank incident and others, were solved mostly by local specialists and in some cases with the help of IAEA and experts from abroad. Installation of digital control console was a big step to strengthen the operational safety. The purpose of this paper is to present 35 years of operating experience of the sole reactor in Bangladesh. Maintenance experience, some modification and upgradation work are also presented through this work.
Keywords: BTRR; N-16 decay tank; SAR; TRIGA Mark-II reactor.