Preliminary Design Studies of a Cylindrical Experimental Hybrid Blanket with Deuterium-Tritium Driver

Şahi̇n, Sümer; Al-Kusayer, Tawfik A.; Raoof, M. A.

doi:10.13182/fst86-a24749

Cited by 83 publications

(5 citation statements)

References 9 publications

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“…A hybrid reactor is a multi-purpose reactor that combines fusion and fission processes. The basic principle in this type of reactor is to convert the fertile materials (U 238 or Th 232 ), located in the blanket surrounding the fusion plasma, into fissile materials (Pu 239 or U 233 ) by transmutation through the capture of the high yield fusion neutrons (Berwald et al 1982;Greenspan 1984;Lee et al 1982;Moir 1982;Raghep et al 1979;Ş ahin & Al-Kusayer 1985;Ş ahin et al 1986, 1991a, 1999, 2001, 2003Ş ahin & Yapici 1999;Ş ahin &Übeyli 2004;Übeyli 2003;Yapıcıet al 2002Yapıcıet al , 2003Yapıcı&Übeyli 2004). In addition to neutron capture process, they may also undergo a substantial amount of fission under the irradiation by high energetic 14 MeV-(D,T) neutrons to increase fusion energy.…”

Section: Introductionmentioning

confidence: 99%

Neutronic analysis of a high power density hybrid reactor using innovative coolants

Yalçın

Übeylı

Acır

2005

Sadhana

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In this study, neutronic investigation of a deuterium-tritium (DT) driven hybrid reactor using ceramic uranium fuels, namely UC, UO 2 or UN under a high neutron wall load (NWL) of 10 MW/m 2 at the first wall is conducted over a period of 24 months for fissile fuel breeding for light water reactors (LWRs). New substances, namely, Flinabe or Li 20 Sn 80 are used as coolants in the fuel zone to facilitate heat transfer out of the blanket. Natural lithium is also utilized for comparison to these two innovative coolants. Neutron transport calculations are performed on a simple experimental hybrid blanket with cylindrical geometry with the help of the SCALE 4•3 System by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and an S 8-P 3 approximation. The investigated blanket using Flinabe or Li 20 Sn 80 shows better fissile fuel breeding and fuel enrichment characteristics compared to that with natural lithium which shows that these two innovative coolants can be used in hybrid reactors for higher fissile fuel breeding performance. Furthermore, using a high NWL of 10 MW/m 2 at the first wall of the investigated blanket can decrease the time for fuel rods to reach the level for charging in LWRs.

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Section: Introductionmentioning

confidence: 99%

Neutronic analysis of a high power density hybrid reactor using innovative coolants

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Übeylı

Acır

2005

Sadhana

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“…Determining the average neutron energy throughout blanket enables us to better understanding the changes in the neutron spectrum. In this study, average neutron energy has been calculated by using below formula taken from earlier work …”

Section: Numerical Results and Discussionmentioning

confidence: 99%

“…In this study, average neutron energy has been calculated by using below formula taken from earlier work. 35…”

Section: Neutron Spectrummentioning

confidence: 99%

Evaluation of the radiation damage parameters of ODS steel alloys in the first wall of deuterium-tritium fusion-fission (hybrid) reactors

Tunç

Şahi̇n

2017

Int J Energy Res

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Summary One of the most important parameters in the design of the fusion‐fission hybrid reactor is the selection of the first wall material. Because the oxide dispersion‐strengthened (ODS) steel alloys have high temperature oxidation, high radiation resistance, good hardness, and corrosion resistance properties, they are thought first wall candidate materials for fusion and fission applications. The objective of this paper is to determine the best radiation damage parameters of various experimental and commercial ODS steels (namely, 12Y1, 12YWT, 1DS, IDK, Eurofer97, MA956, MA957, and PM 2000). Neutron spectrum and average neutron energy throughout blanket, displacement per atom, hydrogen and helium production, nuclear heating, and tritium breeding ratio were calculated by using Monte Carlo methods with Monte Carlo neutron‐photon transport code and nuclear libraries named as ENDF/B‐VI and CLAW‐IV. It is assumed that calculated reactor has been operated full power during a year and neutron wall load is 2.25 MW/m2 (1014 n/s). All investigated first wall materials should be replaced between 3.5 and 4 years. All investigated materials provide minimum required tritium breeding ratio value, and when considering all the calculations performed in this work, 1DS ODS steel is the most suitable first wall materials with respect to other investigated ODS steels.

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“…Spherical blanket, covering the fusion chamber by 4 π simulates the neutron spectrum as in a fusion power plant to a high degree. A closed geometry must be used in order to consider the neutron back and forth scattering between blanket walls correctly and to calculate the authentic fusion neutron spectrum 26,27 …”

Section: Numerical Calculationsmentioning

confidence: 99%

Study on the fusion reactor performance with different materials and nuclear waste actinides

Şahi̇n

Sahiner

et al. 2020

Int J Energy Res

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Summary In this study, both pure fusion blanket and fusion‐fission (hybrid) reactor blanket performance were investigated and discussed separately in two phases. In the first phase, a Monte Carlo radiation damage analysis has been performed for stainless steel (SS304, SS316, and oxide dispersion strengthened (ODS)), molybdenum, vanadium, and tungsten as the first wall (FW) materials, in combination with selected tritium breeders. The main technical parameters for fusion reactors, such as tritium breeding ratio, fusion energy multiplication factor (M), displacement per atom (DPA), and gas production (He, H) have been evaluated. All numerical calculations have been carried out in spherical geometry with MCNP6 code package using continuous energy cross‐sections from the ENDF/B‐VIII.0 library, except DPA calculations. Instead of the ENDF/B‐VIII.0 library, the 30‐group CLAW‐IV library was employed for DPA calculations. Structural material selection for the FW respect to radiation damage limits and reactor performance for energy production and tritium has been concluded. Conventional thermal reactors, such as light water reactors andCanada Deuterium Uranium (CANDU) reactors are producing substantial quantities of transuranic elements, which represent serious nuisance and permanent hazard potential. On the other hand, they become fissionable material under high energetic fusion neutron irradiation and multiply the fusion energy. In the second phase, the investigations are extended to the incineration of minor actinides (MA) in the fusion‐fission (hybrid) mode. The transmutation history of MA nuclear waste is included. MA are added into the first zone of the coolant in TRi‐structural ISOtropic particle TRISO particles with a volume fraction of 6%. The transformation scenario for all MA by SS 304 steel FW is practically the same as with the ODS FW.

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Preliminary Design Studies of a Cylindrical Experimental Hybrid Blanket with Deuterium-Tritium Driver

Cited by 83 publications

References 9 publications

Neutronic analysis of a high power density hybrid reactor using innovative coolants

Neutronic analysis of a high power density hybrid reactor using innovative coolants

Evaluation of the radiation damage parameters of ODS steel alloys in the first wall of deuterium-tritium fusion-fission (hybrid) reactors

Study on the fusion reactor performance with different materials and nuclear waste actinides

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