@article{lee_tayfur_hamza_alzahrani_diaconeasa_2023, title={A Limited-Scope Probabilistic Risk Assessment Study to Risk-Inform the Design of a Fuel Storage System for Spent Pebble-Filled Dry Casks}, url={https://doi.org/10.3390/eng4020094}, DOI={10.3390/eng4020094}, abstractNote={This limited-scope study demonstrates the application of probabilistic risk assessment (PRA) methodologies to a spent fuel storage system for spent pebble-filled dry cask with a focus only on the necessary PRA technical elements sufficient to risk-inform the spent fuel storage system design. A dropping canister scenario in a silo of the spent fuel storage system is analyzed through an initiating event (IE) identification from the Master Logic Diagram (MLD); event sequence analysis (ES) by establishing the event tree; data analysis (DA) for event sequence quantification (ESQ) with uncertainty quantification; mechanistic source term (MST) analysis by using ORIGEN; radiological consequence analysis (RC) by deploying MicroShield, and risk integration (RI) by showing the Frequency-Consequence (F-C) target curve in the emergency area boundary (EAB). Additionally, a sensitivity study is conducted using the ordinary least square (OLS) regression method to assess the impact of variables such as failed pebble numbers, their location in the canister, and building wall thickness. Furthermore, the release categories grouped from the end states in the event tree are verified as safety cases through the F-C curve. This study highlights the implementation of PRA elements in a logical and structured manner, using appropriate methodologies and computational tools, thereby showing how to risk-inform the design of a dry cask system for storing spent pebble-filled fuel.}, journal={Eng}, author={Lee, Joomyung and Tayfur, Havva and Hamza, Mostafa M. and Alzahrani, Yahya A. and Diaconeasa, Mihai A.}, year={2023}, month={Jun} }
 @article{mehboob_al-zahrani_alhusawai_ali_2023, title={Neutronic Analysis of SMART Reactor Core for (U-Th)O-2 and MOX Fuel Hybrid Configurations}, volume={1}, ISSN={["2191-4281"]}, url={http://dx.doi.org/10.1007/s13369-022-07541-7}, DOI={10.1007/s13369-022-07541-7}, journal={ARABIAN JOURNAL FOR SCIENCE AND ENGINEERING}, publisher={Springer Science and Business Media LLC}, author={Mehboob, Khurram and Al-Zahrani, Yahya A. and Alhusawai, Abdulsalam and Ali, Majid}, year={2023}, month={Jan} }
 @article{farag_alzahrani_diaconeasa_2023, title={Pool inlet LOCA safety analysis in support of the emergency core spray system success criteria development of the PULSTAR research reactor}, volume={403}, ISSN={["1872-759X"]}, url={https://doi.org/10.1016/j.nucengdes.2023.112163}, DOI={10.1016/j.nucengdes.2023.112163}, abstractNote={The PULSTAR pool-type research reactor has been operating at power levels up to 1 MWth since initial criticality in 1972 on the North Carolina State University (NCSU) campus. At the current power level, there is no need for an emergency core cooling system to provide additional cooling during abnormal conditions since, at this power level, the peak cladding temperature (PCT) cannot reach the maximum allowable temperature. The range of experiments possible could be extended if the reactor is to be licensed by the U.S. Nuclear Regulatory Commission (NRC) to run up to higher power levels. However, the maximum allowable PCT at higher power levels could be exceeded during abnormal conditions. In this study, a best estimate transient simulation model is used to inform the design of an emergency core spray system to the PULSTAR research reactor. We provide the analysis results of the most limiting scenario of a pool inlet large break loss of coolant accident (LOCA) while operating at higher power levels. Currently, NCSU’s research reactor program is developing a plan to increase the operating power to 2 MWth, thus most of the analysis performed in this paper is prepared at 2 MWth. In addition, a parametric study was carried out to obtain the maximum achievable operating power by having the emergency core spray system installed with the current coolant system arrangement.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Farag, Asmaa S. and Alzahrani, Yahya A. and Diaconeasa, Mihai A.}, year={2023}, month={Mar} }
 @article{farag_alzahrani_diaconeasa_2023, title={Pool inlet LOCA safety analysis in support of the emergency core spray system success criteria development of the PULSTAR research reactor (vol 403, 112163, 2023)}, volume={408}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2023.112342}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Farag, Asmaa S. and Alzahrani, Yahya A. and Diaconeasa, Mihai A.}, year={2023}, month={Jul} }
 @article{al-zahrani_mehboob_alshahrani_abolaban_younis_2022, title={Analysis of SMART reactor core with uranium mononitride for prolonged fuel cycle using OpenMC}, volume={87}, url={http://dx.doi.org/10.1515/kern-2021-1000}, DOI={10.1515/kern-2021-1000}, abstractNote={Abstract}, number={1}, journal={Kerntechnik}, publisher={Walter de Gruyter GmbH}, author={Al-Zahrani, Yahya A. and Mehboob, Khurram and Alshahrani, Tariq F. and Abolaban, Fouad A. and Younis, Hannan}, year={2022}, month={Feb}, pages={82–90} }
 @article{mehboob_al-zahrani_2022, title={Analysis of activated colloidal crud in advanced and modular reactor under pump coastdown with kinetic corrosion}, volume={54}, ISSN={["1738-5733"]}, url={http://dx.doi.org/10.1016/j.net.2022.08.017}, DOI={10.1016/j.net.2022.08.017}, abstractNote={The analysis of rapid flow transients in Reactor Coolant Pumps (RCP) is essential for a reactor safety study. An accurate and precise analysis of the RCP coastdown is necessary for the reactor design. The coastdown of RCP affects the coolant temperature and the colloidal crud in the primary coolant. A realistic and kinetic model has been used to investigate the behavior of activated colloidal crud in the primary coolant and steam generator that solves the pump speed analytically. The analytic solution of the non-dimensional flow rate has been determined by the energy ratio β. The kinetic energy of the coolant fluid and the kinetic energy stored in the rotating parts of a pump are two essential parameters in the form of β. Under normal operation, the pump's speed and moment of inertia are constant. However, in a coastdown situation, kinetic damping in the interval has been implemented. A dynamic model ACCP-SMART has been developed for System Integrated Modular and Advanced Reactor (SMART) to investigate the corrosion due to activated colloidal crud. The Fickian diffusion model has been implemented as the reference corrosion model for the constituent component of the primary loop of the SMART reactor. The activated colloidal crud activity in the primary coolant and steam generator of the SMART reactor has been studied for different equilibrium corrosion rates, linear increase in corrosion rate, and dynamic RCP coastdown situation energy ratio β. The coolant specific activity of SMART reactor equilibrium corrosion (4.0 mg s−1) has been found 9.63 × 10−3 μCi cm−3, 3.53 × 10−3 μC cm−3, 2.39 × 10−2 μC cm−3, 8.10 × 10−3 μC cm−3, 6.77 × 10−3 μC cm−3, 4.95 × 10−4 μC cm−3, 1.19 × 10−3 μC cm−3, and 7.87 × 10−4 μC cm−3 for 24Na, 54Mn, 56Mn, 59Fe, 58Co, 60Co, 99Mo, and 51Cr which are 14.95%, 5.48%, 37.08%, 12.57%, 10.51%, 0.77%, 18.50%, and 0.12% respectively. For linear and exponential coastdown with a constant corrosion rate, the total coolant and steam generator activity approaches a higher saturation value than the normal values. The coolant and steam generator activity changes considerably with kinetic corrosion rate, equilibrium corrosion, growth of corrosion rate (ΔC/Δt), and RCP coastdown situations. The effect of the RCP coastdown on the specific activity of the steam generators is smeared by linearly rising corrosion rates, equilibrium corrosion, and rapid coasting down of the RCP. However, the time taken to reach the saturation activity is also influenced by the slope of corrosion rate, coastdown situation, equilibrium corrosion rate, and energy ratio β.}, number={12}, journal={NUCLEAR ENGINEERING AND TECHNOLOGY}, publisher={Elsevier BV}, author={Mehboob, Khurram and Al-Zahrani, Yahya A.}, year={2022}, month={Dec}, pages={4571–4584} }
 @article{alzahrani_mehboob_abolaban_younis_2021, title={Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO<sub>2</sub>, MOX and (Th/U)O<sub>2</sub> using OpenMC}, volume={86}, url={http://dx.doi.org/10.1515/kern-2020-0063}, DOI={10.1515/kern-2020-0063}, abstractNote={Abstract}, number={3}, journal={Kerntechnik}, publisher={Walter de Gruyter GmbH}, author={Alzahrani, Y. and Mehboob, K. and Abolaban, F. A. and Younis, H.}, year={2021}, month={Jun}, pages={229–235} }
 @article{al-zahrani_mehboob_mohamad_alhawsawi_abolaban_2021, title={Neutronic performance of fully ceramic microencapsulated of uranium oxycarbide and uranium nitride composite fuel in SMR}, volume={155}, url={http://dx.doi.org/10.1016/j.anucene.2021.108152}, DOI={10.1016/j.anucene.2021.108152}, abstractNote={The existing commercial nuclear power plants (PWR and BWR) utilize the oxide fuels, i.e., UO2. This fuel selection is not questionable, where the safety and economy are the top priority in the nuclear industry. In this work, the potential advantages of microencapsulated fuels to System Integrated Modular Advanced Reactor “SMART” has been explored. The UN and UCO have been considered as the candidate fuel material for the SMART reactor. The detailed design and fuel assembly configurations with different fuel types of the SMART reactor have been investigated for criticality with depletion (burn up), fuel and moderator temperature coefficients, and power peaking factor using OpenMC. The fissile and fertile fuel elements with coated particles are embedded in graphite matrices. The results are presented with a review of attributes and potential benefits. The nitride fuel has an advantage of mechanical stability, enhanced thermal conductivity, and high fuel density compared to dioxide fuel (UO2). The melting point of standard fuel, UN, and UCO are similar. Thus, they have higher safety margins under NPP's operating conditions. Therefore, nitride (UN) and carbide (UCO) fuel types are more attractive than standard uranium oxides because they are safer and beneficial. The dioxide fuel (UO2) is considered as the reference and compared with candidate material. The neutronic assessment within SMART core specification examined the effective multiplication factor, thermal flux distribution, axial and radial power distribution, and power peaking factor at the beginning, and end of the fuel cycle.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Al-Zahrani, Yahya A. and Mehboob, Khurram and Mohamad, Daud and Alhawsawi, Abdulsalam and Abolaban, Fouad A.}, year={2021}, month={Jun}, pages={108152} }
 @article{mehboob_al-zahrani_aljohani_alhawsawi_2021, title={Primary coolant activity of 54Mn, 59Fe, 58Co, 60Co, and 51Cr in system integrated small and modular reactor}, volume={160}, url={http://dx.doi.org/10.1016/j.anucene.2021.108356}, DOI={10.1016/j.anucene.2021.108356}, abstractNote={• Analysis of primary coolant activity in SMART reactor. • Use OpenMC as Subroutine in MATLAB. • RCP coastdown model in terms of energy ratio β. • Response of primary coolant activity under dynamic grid load following. • Response of primary coolant activity under limited coolant coastdown. The corrosion results in not only material degradation but also causes CRUD generation in the primary coolant. The axial offset anomaly (AOA) or CRUD Induce Power Shift (CIPS) is caused by CRUD accumulation on the fuel assemblies. It also causes an increase in Occupational Radiation Exposure (ORE). Therefore, in this work, activated CRUD behavior due to 54 Mn, 59 Fe, 58 Co, 60 Co, and 51 Cr has been studied under primary coolant flow transient and power perturbations in a System Integrated Advanced and Modular Reactor (SMART). The “Activated CRUD particles” ACP-SMART code is used to study the primary coolant activity under parabolic corrosion rate. The SMART reactor core has been modeled in OpenMC, and respective group fluxes have been generated. The ACP-SMART, coupled with OpenMC, is modeled and implemented in MATLAB. A dynamic grid load following model and a kinetic coastdown model is implemented in ACP-SMART in terms of the ratio of the coolant kinetic energy and the kinetic energy stored in the rotating parts of a pump (β). The response of 54 Mn, 59 Fe, 58 Co, 60 Co, and 51 Cr activity in primary coolant due to linear, monotonically, and asymptotically coastdown has been investigated. The 59 Fe and 54 Mn showed higher sensitivity relative to the 58 Co, in primary coolant activity corresponding to the dynamic grid load following. However, the 59 Fe and 58 Co activities are more sensitive to coastdown. Nearly all the radioisotopes showed similar behavior under monotonic flow perturbation. 58 Co and 54 Mn have shown a catastrophic behavior in the primary coolant activity under asymptotically coastdown.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Mehboob, Khurram and Al-Zahrani, Yahya A. and Aljohani, Mohammad S. and Alhawsawi, Abdulsalam}, year={2021}, month={Sep}, pages={108356} }
 @article{mehboob_al-zahrani_alhawsawi_2021, title={Primary coolant source term of System-Integrated Modular Advanced Reactor (SMART) due to activated colloidal crud under dynamic grid load following}, volume={384}, url={http://dx.doi.org/10.1016/j.nucengdes.2021.111480}, DOI={10.1016/j.nucengdes.2021.111480}, abstractNote={• Development of activated colloidal crud product in integrated modular advanced reactor “ACCP-SMART” code. • SMART core design in OpenMC. • Colloidal crud behavior under grid load following. • Behaviour of colloidal crud in primary coolant, on core scale and in steam generator tubes. • Colloidal crud response after reactor scram with reactor pumps cost down. The grid load following result in power modulation accelerates the corrosion and activated colloidal crud activity in the primary coolant of PWRs. This paper seeks the response of activated colloidal crud in primary coolant under grid load following resulting in dynamic power modulation. Therefore, a computer code ACCP-SMART “Activated Colloidal Crud Particle in System Integration Modular and Advanced” has been developed. The core of the SMART reactor is modeled in OpenMC to generate the group fluxes under the reactor design constraints. The dynamic grid load following has been employed under the guideline and regulations of IAEA and NEA. The primary coolant activity due to activated colloidal crud has been observed as 51 Cr > 56 Mn > 99 Mo > 24 Na > 59 Fe > 58 Co > 54 Mn > 60 Co. The activated colloidal crud with shorter half-lives ( 24 Na, 54 Mn, and 65 Mn) responded swiftly corresponding to the power modulation under grid load fluctuation. However, the activated colloidal crud with longer half-lives ( 51 Cr, 58 Co, 59 Fe) gradually responds to operated power maneuvering corresponding to the grid load and takes several hours (≈200 h) to attain a new saturation value after the stabilization of reactor operating power. The specific primary coolant activity due to colloidal crud is predominated in the primary coolant. Later, the steam generator activity dominated the primary coolant activity due to the gradual deposition of crud on steam generator tubes. The primary coolant specific activity due to 56 Mn, 54 Mn, 24 Na, and 59 Fe abruptly responded to reactor power modulation in primary coolant and on the core scale. while their response is trivial on steam generator surfaces. On the other hand, the primary coolant specific activity due to activated colloidal crud containing 51 Cr, 58 Co, and 60 Co are less vulnerable and gradually respond to the power modulation and take several hours to attain a new saturation value after power stabilization.}, journal={Nuclear Engineering and Design}, publisher={Elsevier BV}, author={Mehboob, Khurram and Al-Zahrani, Yahya A. and Alhawsawi, Abdulsalam}, year={2021}, month={Dec}, pages={111480} }
 @article{mehboob_al-zahrani_mohamad_2021, title={Simulation of activated corrosion product activity in Korean design system-integrated modular advanced reactor (SMART) under steady-state flow and linearly accelerated corrosion}, volume={134}, url={http://dx.doi.org/10.1016/j.pnucene.2021.103667}, DOI={10.1016/j.pnucene.2021.103667}, abstractNote={The activated corrosion products are the leading source of radiation in Nuclear Power Plants (NPPs). Particularly in advanced integrated systems with a reduction in piping and advanced safety features such as system-integrated modular advanced reactor (SMART), the corrosion product activity is needed to be investigated. In this work, simulation of corrosion product activity has been evaluated for the SMART reactor under normal and shutdown conditions and compared with a typical Pressurized Water Reactor (PWR). For this purpose, a computer code ACP-SMART has been developed and implemented in MATLAB, which uses OpenMC as a subroutine. The effective group cross-section has been generated by OpenMC under the core design constraints. The results for 56Mn, 59Fe, 24Na, 99Mo, 58Co, and 60Co have been found as 0.072 μCi/cm3, 0.240 μCi/cm3, 0.290 μCi/cm3, 0.360 μCi/cm3, 0.015 μCi/cm3, and 0.020 μCi/cm3, respectively. The total specific activity due to corrosion products in the core, primary coolant, and inside the steam generator is 2.790 μCi/cm3, 0.183 μCi/cm3, and 0.145 μCi/cm3, respectively, which are 89.5%, 5.8%, and 4.7% of the overall contribution. Results indicate that the primary coolant and steam generator specific activity reach their saturation values fairly rapidly. Predominant corrosion product activity during normal operation is due to 56Mn while 58Co and 59Fe dominate after reactor shutdown. 99Mo remains dominant to 24Na and 59Fe during normal operation while 59Fe leads over 99Mo and 24Na after shutdown. A linearly increasing corrosion rate has been employed, and an effect on saturation activity has been investigated. For the linearly increasing corrosion rate, the specific activity behavior has changed considerably. The time taken to reach the saturation activity strongly depends upon the corrosion rate. The saturation activity in the primary coolant and steam generator depends on the CRUD removal rate from the core scale.}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Mehboob, Khurram and Al-Zahrani, Yahya A. and Mohamad, Daud}, year={2021}, month={Apr}, pages={103667} }
 @article{mehboob_alzahrani_fallatah_qutub_younis_2020, title={Radioactivity and Radiation Hazard Indices Assessment for Phosphate Rock Samples from Al-Jalamid, Turaif, Umm Wu’al, and As-Sanam, Saudi Arabia}, volume={9}, url={http://dx.doi.org/10.1007/s13369-020-04929-1}, DOI={10.1007/s13369-020-04929-1}, journal={Arabian Journal for Science and Engineering}, publisher={Springer Science and Business Media LLC}, author={Mehboob, Khurram and Alzahrani, Yahya A. and Fallatah, O. and Qutub, M. M. T. and Younis, Hannan}, year={2020}, month={Sep} }