Şahin, Sümer
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Sahin, Suemer
S.,Sumer
Sahin, Sumer
S., Sahin
S., Sumer
S.,Sahin
Sumer, Sahin
S.,Şahin
Ş.,Sümer
Sahin,S.
Şahin, Sümer
Şahin,S.
Sümer, Şahin
S.,Sumer
Sahin, Sumer
S., Sahin
S., Sumer
S.,Sahin
Sumer, Sahin
S.,Şahin
Ş.,Sümer
Sahin,S.
Şahin, Sümer
Şahin,S.
Sümer, Şahin
Job Title
Profesör Doktor
Email Address
Main Affiliation
Department of Mechanical Engineering
Status
Former Staff
Website
ORCID ID
Scopus Author ID
Turkish CoHE Profile ID
Google Scholar ID
WoS Researcher ID
Sustainable Development Goals
14
LIFE BELOW WATER
0
Research Products
2
ZERO HUNGER
0
Research Products
11
SUSTAINABLE CITIES AND COMMUNITIES
2
Research Products
1
NO POVERTY
0
Research Products
12
RESPONSIBLE CONSUMPTION AND PRODUCTION
1
Research Products
7
AFFORDABLE AND CLEAN ENERGY
18
Research Products
5
GENDER EQUALITY
0
Research Products
3
GOOD HEALTH AND WELL-BEING
3
Research Products
9
INDUSTRY, INNOVATION AND INFRASTRUCTURE
0
Research Products
13
CLIMATE ACTION
0
Research Products
6
CLEAN WATER AND SANITATION
5
Research Products
10
REDUCED INEQUALITIES
0
Research Products
4
QUALITY EDUCATION
0
Research Products
15
LIFE ON LAND
0
Research Products
16
PEACE, JUSTICE AND STRONG INSTITUTIONS
0
Research Products
17
PARTNERSHIPS FOR THE GOALS
0
Research Products
8
DECENT WORK AND ECONOMIC GROWTH
2
Research Products
This researcher does not have a Scopus ID.
This researcher does not have a WoS ID.
Scholarly Output
45
Articles
24
Views / Downloads
12/0
Supervised MSc Theses
0
Supervised PhD Theses
0
WoS Citation Count
266
Scopus Citation Count
350
WoS h-index
10
Scopus h-index
12
Patents
0
Projects
0
WoS Citations per Publication
5.91
Scopus Citations per Publication
7.78
Open Access Source
4
Supervised Theses
0
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| Journal | Count |
|---|---|
| Energy Conversion and Management | 8 |
| Progress in Nuclear Energy | 4 |
| 15th International Conference on Emerging Nuclear Energy Systems -- MAY 15-19, 2011 -- San Francisco, CA | 3 |
| International Journal of Energy Research | 3 |
| International Journal of Hydrogen Energy | 3 |
Current Page: 1 / 4
Scopus Quartile Distribution
Competency Cloud
45 results
Scholarly Output Search Results
Now showing 1 - 10 of 45
Editorial Preface To the Special Issue on "17th International Conference on Emerging Nuclear Energy Systems (icenes'2015), 4-8 October 2015, Istanbul, Turkey"(Pergamon-elsevier Science Ltd, 2016) Sahin, Sumer; Sahin, Haci Mehmet; Martinez-Val, Jose; Wu, Yican[No Abstract Available]Editorial Citation - WoS: 1Citation - Scopus: 1Editor's Notes on Icenes'2013, 16th International Conference on Emerging Nuclear Energy Systems(Pergamon-elsevier Science Ltd, 2015) Sahin, Sumer[No Abstract Available]Article Citation - WoS: 7Citation - Scopus: 10Evaluation of Integral Quantities in an Accelerator Driven System Using Different Nuclear Models Implemented in the Mcnpx Monte Carlo Transport Code(Pergamon-elsevier Science Ltd, 2013) Sarer, Basar; Sahin, Sumer; Celik, Yurdunaz; Gunay, MehtapThe MCNPX code offers options based on physics packages; the Bertini, ISABEL, INCL4 intra-nuclear models, and Dresner, ABLA evaporation-fission models and CEM2k cascade-exciton model. This study analyzes the main quantities determining ADS performance, such as neutron yield, neutron leakage spectra, heating and neutron and proton spectra in the target and in the beam window calculated by the MCNPX-2.5.0 Monte Carlo transport code, which is a combination of LAHET and MCNP codes. The results obtained by simulating different models cited above and implemented in MCNPX are compared with each other. The investigated system is composed of a natural lead cylindrical target and stainless steel (HT9) beam window. The target has been optimized to produce maximum number of neutrons with a radius of 20 cm and 70 cm of height. The target is bombarded with a high intensity linear accelerator by a 1 GeV, 1 mA proton beam. The protons are assumed uniformly distributed across the beam of radius 3 cm, and entering the target through a hole of 5.3 cm radius. The proton beam has an outer radius of 53 cm and an inner radius of 5.0 cm. The maximum value of the neutron flux in the target is observed on the axis similar to 10 cm below the beam window, where the maximum difference between 7 different models is similar to 15%. The total neutron leakage of the target calculated with the Bertini/ABLA is 1.83 x 10(17) n/s, and is about 14% higher than the value calculated by the INCL4/Dresner (1.60 x 10(17) n/s). Bertini/ABLA calculates top, bottom and side neutron leakage fractions as 20%, 2.3%, 77.6% of the total leakage, respectively, whereas, the calculated fractions are 18.6%, 2.3%, 79.4%, respectively, with INCL4/Dresner combination. The largest heat deposition density by considering all particles in the beam window calculated with CEM2k model is 104 W/cm(3)/mA, which is 9.0% greater than the lowest value predicted with INCL4/Dresner model (95.4 W/cm(3)/mA). The maximum average heat deposition density for all particles in the target is calculated as 6.87 W/cm(3)/mA with INCL4/ABLA. (C) 2013 Elsevier Ltd. All rights reserved.Article Citation - WoS: 3Citation - Scopus: 3Incidence of Medical Device-Related Pressure Injuries and Identification of Risk Factors in the Neonatal Unit(Elsevier Sci Ltd, 2024) Yarkiner, Zalihe; Bahar, Arzu; Sonmez, Munevver; Kapan, Emine; Sahin, Simge; Kostekci, Ezgi; Erdeve, OmerAim: This study was conducted to investigate the incidence of medical device-related pressure injuries (MDRPIs) and the risk factors influencing their occurrence in the neonatal intensive care unit (NICU). Method: This study is a prospective, descriptive study. The research was conducted with 116 newborns between June 1, 2022, and June 1, 2023. Newborns who stayed in the neonatal intensive care unit for at least 24 h were observed daily for medical device-related pressure injuries under and around each medical device throughout their stay in the intensive care unit. The "Case Report Form," "MDRPIs Monitoring Form," "Braden Q scale for children," National Pressure Injury Advisory Panel (NPIAP) Pressure Grading, and Glasgow Coma Scale were used in the research. Results: The incidence of medical device-related pressure injuries is 35.3 % (41/116). It was found that 38.1 % (16/42) of medical device-related pressure injuries developed due to Near-Infrared Spectroscopy (NIRS) probes, and 33.5 % (14/42) developed due to medical devices related to the respiratory system. In terms of anatomical location, 38.1 % occurred on the forehead, and 23.8 % on the arm/leg. The difference between birth weight, gestational age, development of MDRPIs in newborns receiving sedation and inotropes was found to be statistically significant. Regression analysis identified gestational age (p = 0.040, OR = 0.795, 95%CI = [0.632-1.000]) as an independent risk factor for the occurrence of medical device-related pressure injuries. Conclusions: The incidence of medical device-related pressure injuries in newborns was relatively high in this study, with gestational age being the most significant risk factor for MDRPIs formation. It is crucial for neonatal intensive care nurses to consider associated risk factors while providing newborn care and implement appropriate preventive measures to reduce the incidence of MDRPIs.Conference Object Citation - WoS: 9Citation - Scopus: 10Comparisons of the Calculations Using Different Codes Implemented in Mcnpx Monte Carlo Transport Code for Accelerator Driven System Target(Amer Nuclear Soc, 2012) Sarer, Basar; Sahin, Sumer; Gunay, Mehtap; Celik, YurdunazThe MCNPX code offers options based on physics packages; the Bertini, ISABEL, INCL4 intra-nuclear models, and Dresner, ABLA evaporation-fission models and CEM2k cascade-exciton model. The study analyzes the main quantities determining ADS performance such as neutron yield, neutron leakage spectra, and neutron and proton spectra in the target and in the beam window calculated by the MCNPX-2.5.0 Monte Carlo transport code, which is a combination of LAHET and MCNP codes. The results obtained by simulating different models, cited above and implemented in MCNPX are compared with each other. The investigated system is composed of a natural lead cylindrical target and stainless steel (HT9) beam window. Target has been optimized to produce maximum number of neutrons with a radius of 20 cm and 70 cm of height. Target is bombarded with a high intensity linear accelerator by a 1 GeV, 1 mA proton beam. The protons are assumed uniformly distributed across the beam of radius 3 cm, and entering the target through a hole of 5.3 cm radius. The proton beam has an outer radius of 5.3 cm and an inner radius 5.0 cm. The maximum of the neutron flux in the target is observed on the axis similar to 10 cm below the beam window, where the maximum difference between 7 different models is similar to 15 %. The total neutron leakage out of the of the target calculated with the Bertini/ABLA is 1.83x10(17) n/s, and is about 14 % higher than the value calculated by the INCL4/Dresner (1.60x10(17) n/s). Bertini/ABLA calculates top, bottom and side neutron leakage fractions as 20 %, 2.3 %, 77.6 % of the total leakage, respectively, whereas, they become 18.6 %, 2.3 %, 79.4 % with INCL4/Dresner combination.Article Citation - WoS: 22Citation - Scopus: 26Hydrogen Hazard and Mitigation Analysis in Pwr Containment(Pergamon-elsevier Science Ltd, 2013) Sahin, Sumer; Sarwar, Mohammad SohailThis paper describes the analytical results for the estimation of hydrogen concentration in the containment atmosphere based on zirconium oxidation reaction following a severe accident. The analysis provides useful information about the potential challenge of local hydrogen accumulation in the containment, which may be used to reduce the hydrogen detonation risk and to design the capacity and arrangements of mitigation measures. The containment analysis has been performed using computer code COGAP which uses the scenario of loss of coolant accident. The behavior of pressure and hydrogen concentrations in containment as a function of time under the severe accident condition is presented in graphical form. The mitigation measures (recombiners) are essential to maintain containment atmosphere in the safe stable conditions. A hydrogen control system is to mitigate the hydrogen risk by comparing results from a reference accident sequence with and without recombiners. This comparison show that combustible gas occur in few local areas in the containment for a limited time span and hydrogen concentration is reduced significantly with the use of recombiners. (C) 2013 Elsevier Ltd. All rights reserved.Article Editorial Notes on the 2012 International Youth Nuclear Congress (iync), Charlotte, North Carolina, Usa (5–11 August 2012)(Energy Conversion and Management, 2012) Şahin, SümerPrimary purpose of the International Youth Nuclear Congress (IYNC) series is to transfer knowledge from the current generation of leading scientists and engineers to the next generation. Scien tific, political, public and corporate views regarding the develop ment of different nuclear issues are presented to provide comprehensive discussions on all sides of the subject. With this aim, The 2012 International Youth Nuclear Congress has been held in Charlotte, North Carolina. IYNC2012 was focused on the use of nuclear energy more than ever after the accident at Fukushima power plant and also in the framework of the nuclear renaissance in many developing countries and world powers as well. IYNC2012 offered the opportunity to share knowledge, experience, best prac tices and information about nuclear energy between generations and also between peers in the ongoing mission to promote the peaceful use of nuclear power.Article Citation - WoS: 3Citation - Scopus: 4Experimental Evaluation of Surveillance Capsule Assemblies for Life Assessment of Chasnupp Unit-1 Reactor Pressure Vessel(Pergamon-elsevier Science Ltd, 2016) Sahin, Sumer; Saeed, AsimNeutron flux and energy spectrum were determined at the surface of three in-vessel Surveillance Capsule Assemblies (SCAs) removed from CHASNUPP Unit-1 after 2nd, 4th, and 9th fuel cycles for the life assessment of reactor pressure vessel belt line region. Dosimetry data were measured from radiometric sensors irradiated in base material section of SCAs. Fast neutron flux (E > 1.0 MeV) was best estimated at the surface of three SCAs corresponding to the center of C-1 core using the least square method by employing LSL-M2 package. These results were compared with fast neutron flux calculated using DOT3.5 code and both results are within good agreement of +/- 20% acceptance criteria as described in Regulatory Guide 1.190. Therefore, calculational model was validated by dosimetry evaluation and these results can be used in the life assessment of CHASNUPP Unit-1 pressure vessel belt line region. (C) 2015 Elsevier Ltd. All rights reserved.Article Citation - WoS: 21Citation - Scopus: 27LIFE hybrid reactor as reactor grade plutonium burner(Pergamon-elsevier Science Ltd, 2012) Sahin, Sumer; Sahin, Haci Mehmet; Acir, AdemThe early version of the conceptual modified design of the Laser Inertial Confinement Fusion Fission Energy (LIFE) engine consists of a spherical fusion chamber of 5 m diameter, surrounded by a multi-layered blanket. The first wall is made of 2 cm thick ODS and followed by a Li17Pb83 zone (2 cm), acting as neutron multiplier, tritium breeding and front coolant zone. It is separated by an ODS layer (2 cm) from the FLIBE molten salt zone (50 cm), containing fissionable fuel. A 3rd ODS layer (2 cm) separates the molten salt zone on the right side from the graphite reflector (30 cm). Calculations have been conducted for a constant fusion driver power of 500 MWth, in S-8-P-3 approximation using 238-neutron groups. Reactor grade (RG) plutonium carbide fuel in form of TRISO particles with volume fractions of 2%, 3%, 4%, 5% and 6% have been dispersed homogenously in the FLIBE coolant. Tritium breeding ratio (TBR) values per incident fusion neutron for the above cited cases start with TBR = 1.35, 1.52, 1.73, 2.02 and 2.47, respectively. With the depletion of fissionable RG-Pu isotopes, TBR decreases gradually. At startup, higher fissionable fuel content in the molten salt leads to higher blanket energy multiplication, namely M-0 = 3.8, 5.5, 7.7, 10.8 and 15.4 with 2%, 3%, 4%, 5% and 6% TRISO volume fraction, respectively. Calculations have led to very high burn up values (>400,000 MD.D/MT). TRISO particles can withstand such high burn ups. Such high burn ups would lead to drastic reduction of final nuclear waste per unit energy production. (C) 2012 Elsevier Ltd. All rights reserved.Conference Object Transmutation of Minor Actinides in Candu Reactors(2010) Şahin,S.; Şahin,H.M.; Acir,A.; Al-Kusayer,T.A.Large quantities of nuclear waste plutonium have been accumulated in the civilian LWRs and CANDU reactors in form of minor actinides (MAs). Reactor grade plutonium and other transuranium elements can be used as a booster fissile fuel material in form of mixed ThO2/MAO2 fuel in a CANDU fuel bundle in order to assure reactor criticality. Following fuel compositions have been selected for investigations; Reactor grade plutonium: Circled digit one 96 % thoria (ThO2) + 4 % PuO2 and Circled digit two 91 % ThO2 + 5 % UO2 + 4 % PuO2. The latter is used for the purpose of denaturing the new 233U fuel with 238U. The behavior of the criticality k∞ and the burn-up values of the reactor have been pursued by full power operation for > ∼ 8 years. The reactor starts with k∞ = ∼ 1.39 and the criticality drops down asymptotically to values k∞ > 1.06, still acceptable and useable in a CANDU reactor. Reactor criticality k ∞ remains nearly constant between the 4th year and 7th year of plant operation and then a slight increase is observed thereafter, along with a continuous depletion of thorium fuel. Totality of nuclear waste actinides after the extraction of uranium isotopes: The best fuel compositions with respect to power flattening as well as long term reactivity have been found by mixing thoria with 14 % minor actinides in form of MAO 2 in the central fuel bundle and decreasing the MAO2 content in radial direction at discrete levels down to 2 % at the periphery. The temporal variation of the criticality k∞ and the burn-up values of the reactor have been calculated for a period of 10 years, operated at full power. The criticality starts at time zero near to k∞ = ∼ 1.24 for both fuel compositions. A sharp decrease of the criticality has been observed during the first year as a consequence of rapid plutonium burnout in the actinide fuel. The criticality becomes quasi constant after the 2 nd year after sufficient 233U is accumulated and remains close to k∞,end = ∼1.06 over ∼ 10 years. Quasi-uniform power generation density has been realized in the fuel bundle throughout the reactor operation. In all investigated cases, plutonium burns up rapidly and after the 2nd year, the CANDU reactor begins to operate practically as a thorium burner.
