International Journal of Nuclear Energy Science and Technology (14 papers in press)
Determination of alpha radioactivity in soil samples collected from University of Kerbala, lraq
by Ali Abojassim
Abstract: In this study, alpha particles in 60 soil samples of the University of Kerbala (Freiha sites) were determined using a CN-85 detector. The results show that the average value of radon concentrations in air space, radon concentrations 222Rn in samples, annual effective dose, radium content, mass exhalation rates, surface exhalation rates, and uranium concentrations were 120.82
Keywords: aqlpha particles; soil samples; CN-85 ; University of Kerbala; radon gas; Iraq.
Evaluation of iodine-129 transmutation fraction in high flux reactors
by Fahim Tighemine, Naima Amrani, Ahmed Boucenna, A. Abdelghafar Galahom
Abstract: This study aims to transmute iodine-129 (I-129) into a stable isotope of xenon-130 (Xe-130) by irradiation in the high flux reactor. I-129 is considered one of the most hazardous long-lived fission products. The numerical results of the transmutation fraction, the transmutation amount of I-129 and mass evolution of Xe-130 produced in three high flux reactors, namely Petten, BR2 and SM3, are simulated for an irradiation time of 300 effective full power days using the ChainSolver 2.34 code. Based on the results obtained, it is recommended to use SM3 as the most effective I-129 transmutation application.
Keywords: transmutation fraction; long-lived fission product; iodine-129; iodine-127; ChainSolver 2.34; high flux reactor.
In-core fuel management of TRIGA reactor optimising performance and safety
by Md. Hossen Altaf, Mohammad Sayem Mahmood
Abstract: The TRIGA Mark II research reactor has been under operation in Bangladesh since 1986 and postponed radioisotope production in 2008. The possibility to extend the length of the operating cycle of the reactor core by compacting the inner region with fuels replacing the graphite elements has been investigated. This leads to the increment of core excess reactivity at the expense of reactor performance. Confirmation of the safety has been demonstrated by departure of nucleate boiling analysis. The result finds promising prospects for the future analysis with explicit thermal hydraulic safety calculation for the compact core configuration of the reactor.
Keywords: MCNP; TRIGA; in-core fuel management.
Quantum mechanical calculations of fusion reactions induced by multi-neutron halo systems below and above the Coulomb barrier
by Fouad A. Majeed, Fatima M. Hussain
Abstract: The role of the breakup channel in fusion reactions involving two and four neutron halos is discussed in this paper. The cross sections ?_fus, barrier distributions D_fus, probability of fusion P_fus , and the mean angular momentum ?L? were calculated using quantum mechanics by means of the CC code. The systems 4He+65Cu, 6He+64Zn and 6He+197Au were used as base systems to study the fusion reaction for the systems 6He+65Cu, 8He+65Cu, 6He+63Cu, and 6He+192Os to understand the effect of the breakup of 6He and 8He halo nuclei. The Woods-Saxon parameters for the studied systems were fitted using the ?^2 method to fit the centroid with the experimental height barrier Vb. The comparison of theoretical results with the corresponding experimental data shows clearly that for these halo systems the breakup channel plays a crucial role in the calculations and should be taken into consideration to enhance the calculations, especially below the Coulomb barrier Vb.
Keywords: fusion cross section; mean angular momentum; fusion barrier distribution; breakup channel.
Detailed safety assessment for the VVER-1000 fuel assembly
by Mohamed Mohsen, Mohamed Abdel-Rahman
Abstract: This paper presents three main studies for one fuel assembly of the VVER-1000. The first study is the neutronic analysis using MCNPX 26f. The second study is the thermal-hydraulic analysis. This second study has been simulated analytically and numerically using MATLAB and COMSOL-Multiphysics physics respectively. The third study is the solid mechanics analysis by using COMSOL-Multiphysics physics. The sequence of the above studies is set as follows. During the neutronic analysis, the main safety-related parameters, fuel burn-up calculations, and the power mapping are simulated. In addition, from the radial power peaking factor (PPF) distribution, the hot channel is determined. This is in order to conduct the thermal hydraulic and solid mechanics analyses. The analytical solution of the thermal-hydraulic revealed that the maximum fuel and clad temperatures are 1398.9 K and 653.95 K, respectively. In addition, the minimum departure from nucleate boiling ratio (MDNBR) is 2.04. On the other hand, the numerical solution of the thermal-hydraulic revealed that the maximum fuel and clad temperatures are 1040.04 K and 625.03 K, respectively, and the MDNBR equals 2.07. Finally, the solid mechanics analysis revealed that the maximum von Mises stresses acting on the fuel and clad materials are 91.81 MPa and 40.82 MPa respectively, and the maximum fuel outer surface displacement equals to 0.06023 mm. The results obtained from this paper are in a good agreement with both the FSAR and the previous published works.
Keywords: VVER-1000; MCNPX; neutronic analysis; thermal-hydraulic analysis; solid mechanics analysis; PPF; hot channel; MATLAB; COMSOL-Multiphysics; MDNBR; FSAR.
Production cross-sections calculations of proton and neutron induced reactions for producing copper radionuclide used in PET imaging and radiation therapy applications
by Nabeel F. Lattoofi, Ali A. Al Zubaidy
Abstract: In the last decades, many radioisotopes have been widely used in medical imaging diagnostic and radiation therapy applications. Among these are the copper isotopes, which are the most important isotopes used in these fields. These isotopes have been produced in reactors with different nuclear reactions in which a non-copper element has been used as a target. In our present research, we aimed to calculate the nuclear (p, n) and (n, p) reactions cross-section used in producing different copper isotopes, namely 60, 61, 62, 63, 64, 67Cu, using the Empire code for this purpose. The calculated results, investigated with different phenomenological and microscopic models for level density, have been compared with the available experimental data taken from the EXFOR database for energy ranging from 0 MeV up to 30 MeV. The results show a good agreement with experimental ones for (p, n) reactions with Hartree Fock Bogoliubov microscopic model (HFBM). However the cross section calculations with Enhanced Generalized Superfluid Model (EGSM) for (n ,p) reactions are closer to the experimental data.
Keywords: level density model; cross-section; therapeutic radioisotope.
Structural response evaluation of reinforced-concrete nuclear containment structure subjected to internal overpressure with high-temperature loading
by Priya Arjunsinh Mahida, Dimple Desai
Abstract: The main aim of this study was to investigate the structural response of a nuclear containment structure subjected to internal overpressure and higher temperature; these conditions, which occur beyond the design basis, can be caused by accidents (such as Loss of Coolant Accident (LOCA)) in a nuclear power plant. To forecast the structural response of a nuclear containment structure under internal overpressure and high-temperature loading conditions, a non-linear analysis was conducted with the finite element analysis software SAP 2000. More specifically, the layered shell elements in SAP 2000 were used for the 3D finite-element modelling of the reinforced-concrete nuclear containment structure. The layered shell elements represent the non-linear behaviour of the concrete and steel, while the reinforcement was modelled with two layers in circumferential and meridional directions. The structural behaviour of the reinforced-concrete containment structure subjected to various loadings (internal overpressure and high-temperature) is presented in terms of maximum leakage rate, the displacement and stress concentration in the containment structure. According to the results, the reinforced-concrete nuclear containment experiences a safety failure at 1.2 Db (Db: design basis internal pressure and temperature) and structural failure at 1.22 Db.
Keywords: LOCA; loss of coolant accident; finite element analysis; nuclear containment structure; high temperature; internal overpressure.
Calculation of fuel burnup and excess reactivity using TRIGLAV code for the BAEC TRIGA research reactor
by Md. Mizanur Rahman, Anisur Rahman, Shafiul Hossain, P.K. Das, Md. Ashraf Ali, Md. Abdul Malek Soner, Abdullah-Al-Mahmud
Abstract: The TRIGLAV computer code, a computation analysis package tool, was used for the first time in BAEC TRIGA research reactor (BTRR) to calculate burnup and excess reactivity. Since its commissioning in 1986, the reactor has been operating around 800 megawatt days for research, education, training and radioisotope production without any reshuffling or reloading of core. In this study, the fuel burnup and excess reactivity for BTRR are calculated using TRIGLAV code. The calculated burnup result of TRIGLAV code is compared with the results of TRIGAP and MVP-BURN code to find out the usefulness of TRIGLAV code and the result is found reasonable. The calculated excess reactivity data is also compared with actual operational data and safety analysis report and were found in good agreement. The individual fuel burnup data can be used to reshuffle the fuel and the excess reactivity data to predict the core life. The calculated total burnup, identification of hottest fuel and excess reactivity data shows that the reactor can be operated safely for another 800 megawatt days regarding the requirement of burnup and excess reactivity.
Keywords: TRIGLAV; BTRR; excess reactivity; burnup; nuclear reactor.
A comparison based on the calculation of some hadron doses in addition to the secondary neutron fluence using the FLUKA Monte Carlo code
by Mohammed Yjjou, Hassane Dekhissi, Jamal Eddine Derkaoui, Abdessamad Didi, Adil Aknouch
Abstract: In this paper, we introduce a new comparison based on the calculation of some hadron doses in addition to the secondary neutron fluence using the FLUKA MC code. At first, the dose depths of the hadrons were compared, which provided the description of the difference between their distributions and those of the photons. After that, it was proposed that a tumour was at a precise depth. Then the energy used only by protons and carbons was tested because of the usefulness they possess at this depth. In a second phase, the secondary neutrons' fluence given by protons and carbon ions was compared according to their energy and solid angle of production. This study shows the importance of taking into account the distribution of secondary particles, especially neutrons, along with the dose distribution of accelerated hadrons in hadrotherapy to improve the irradiation accuracy.
Keywords: Monte Carlo codes; hadrotherapy; dose; hadrons; fluence; secondary neutrons; tumour.
Analysis of a pressurised water reactor-based nuclear accident using PCTRAN simulator and fuzzy expert system
by Altab Hossain, Intisher Al-Tahmid Omi, Miskat Islam Anika, Jannat Mahal
Abstract: Potential malfunctions in nuclear reactor operation may cause undesirable consequences or accident. Hence, it is important to analyse the influence of different malfunctions such as Loss of Coolant Accident (LOCA) in hot leg, loss of AC power, turbine trip and loss of flow. Therefore, a pressurised water reactor (VVER-1200) based accident analysis has been carried out. Furthermore, a risk analysis using Fuzzy Expert System (FES) is conducted. FES comprises two Fuzzy Inference Systems (FIS) such as FIS-1 and FIS-2. The input parameters of FIS-1 are obtained using the Personal Computer Transient Analyser (PCTRAN). Subsequently, FIS-2 involves the relationship between inputs 'severity' (output of FIS-1) and 'frequency of failure rate for operating days' and the final output 'Risk Level'. During LOCA, for inputs Reactor Coolant System Pressure (RCSP) deviation of 69 bar, average Temperature of Fuel (TAF) deviation of 496.7°C and number of control rods inserted after run-time of 5, the outputs severity and risk level are found to be 84.6% and 91.7% respectively. A similar pattern is found during loss of flow. On the other hand, the level of risk for loss of AC power and turbine trip is found to be 30% and 93.6%, respectively.
Keywords: malfunctions; nuclear reactor; accident analysis; loss of coolant accident; FES; risk level; temperature of fuel; control rods; severity; turbine trip.
The current public acceptance in Brazil of nuclear science and technology for peaceful purposes
by Vitor Fernandes De Almeida, Luciana Sampaio Ribeiro, Edilaine Ferreira Da Silva, Anna Flávia De Freitas Valiante Peluso, Nathália Silva De Medeiros, Amir Zacarias Mesquita
Abstract: The development of nuclear science in Brazil in various fields is significant. In the Brazilian scenario, this science was effectively stimulated in the mid-1950s, with the first radiological research centre and nowadays, the country boasts four research reactors and two nuclear power reactors in operation, and 11 fuel cycle facilities. These units possess a nominal production of about 2.000 MWe, which represents 3% of the national matrix. However, this percentage does not reflect the full nuclear energetic potential in Brazil, once the country has the seventh-largest uranium reservoir worldwide and is one of the few countries that has mastered the nuclear fuel cycle. Given the lack of public knowledge about nuclear technology in Brazil, an initiative for assessing Brazilian public comprehension and approval on nuclear science has emerged. In face of the collected and analysed data, it was possible to observe the degree of knowledge about radiopharmaceuticals and nuclear energy.
Keywords: nuclear energy; nuclear technology; radiopharmaceuticals; radiation; survey.
Radiation dosimetry for thyroid cancer patients and validation using 3D printing phantoms
by M.M. Emad El-din, Eman Massoud, Ahmed Yousry El-Agamawi, A.M. Kany, M.R. Ezz El-din
Abstract: This research presents the external radiation exposure resulting from Well Differentiated Thyroid Cancer (WDTC) patients after receiving radioiodine (I-131) activities orally and its decay as a function of time. Dose rates were recorded at different time post-administration using a survey meter for 86 thyroidectomised WDTC patients during follow-up diagnosis procedures and during radiotherapy hospitalisation period. Afterwards, a 3D phantom was designed based on Computed Tomography (CT) images to validate the external exposure. Retained activities were found to be 32.1, 13.7, 6.3 and 3.1% within 24, 48, 72 and 96 h respectively. The measurements acquired from the 3D phantom were 14.3:18.1% higher than real patients. Dose rates were less than 10 µSv/h at 3 m distance that is considered safe distance from patients during diagnosis procedures. The administrated activity decays via bi-exponential relation with time, and it is recommended to quarantine patients undergoing post-operative radiotherapy for 36 hours post-administration at least.
Keywords: radioiodine-131; the external radiation exposure; WDTC dosimetry; 3D printing; radiotheranostics.
Neutronic evaluation of annular fuel assemblies
by Raphael Henrique Martins Silva, Clarysson Alberto Melo Da Silva
Abstract: The concept of annular fuel rods was initially studied in the 1950s, but researchers have expanded the investigations considerably from 2000 to improve the performance of nuclear power plants. Thus, aiming a probable replacement of the fuel assemblies of Angra II reactor, a PWR (Pressurised Water Reactor) type in operation in Brazil, this paper presents a neutronic comparison between the Typical Fuel Assembly (TFA) and the Annular Fuel Assemblies (AFAs). Five different AFA geometries are compared with a TFA to evaluate the influence in the criticality, neutron flux and the Doppler Coefficient of Reactivity. The Monte Carlo N-Particle Transport code, Version 6 (MCNP6), was used to calculate the neutronic parameters. The results present small variations among the simulated fuel assemblies, where the AFA with 11x11 fuel pins presents better agreement in relation to criticality and neutron flux values when compared with the TFA.
Keywords: pressurised water reactor; advanced fuels; annular fuels; neutronic analysis; MCNP6.
Relativistic fictitious forces from the perspective of the accelerated rotating platform
by A. Sfarti
Abstract: In the current paper we present the expression of the relativistic fictitious forces as measured in the frame of a non-uniformly rotating platform. The solution is of great interest for real time applications because Earth-bound laboratories are inertial only in approximation. The motivation is that the real life applications include accelerating and rotating frames more often than the idealised case of inertial frames; our daily experiments happen in the laboratories attached to the rotating Earth. The accelerations play an important role in centrifuges ramping up to speed. We will provide a straightforward method of deriving the fictitious forces arising in the rotating frame in their relativistic form. We are also correcting the expression of the Euler force in its classical (non-relativistic) form by correcting an error in its derivation that has persisted for centuries. This paper is dedicated to professors Mikayel Sarian and Eliza Haseganu.
Keywords: general coordinate transformations; uniform rotation; relativistic Coriolis force; relativistic centrifugal force; relativistic Euler force.