Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters in (D-T) driven hybrid reactor system

Annals of Nuclear Energy

Volumn 28, Issue 18, 2001, Pages 1825-1850

  Yapici, Hüseyin ^a   Ipek, Osman ^b   Özceyhan, Veysel ^a  

^a ERCIYES UNIVERSITY   (Turkey)

^b SULEYMAN DEMIREL UNIVERSITY   (Turkey)

Author keywords

[No Author keywords available]

Indexed keywords

COOLANTS; DEUTERIUM; FISSION PRODUCTS; LITHIUM; MELTING; TEMPERATURE DISTRIBUTION; TRITIUM;

NEUTRONIC PERFORMANCE; NUCLEAR FUEL RODS;

NUCLEAR REACTORS;

TRITIUM;

EID: 0035546628     PISSN: 03064549     EISSN: None     Source Type: Journal    
DOI: 10.1016/S0306-4549(01)00020-2     Document Type: Article

References (41)

1. Tritium breeding in fusion
2. Blanket design and calculated performance for the LOTUS fusion-fission hybrid test facility
3. Basic requirements for A 1000-MW(electric) class Tokamak fusion-fission hybrid reactor and its blanket concept
4. Properties for reactor safety analysis
5. Fusion-fission hybrid reactor
6. Power density flattening in fusion-fission hybrid reactors
7. SCDAP/RELAP%/MOD2 code manual
8. 4 for various coolants to breed fissile fuel for LWRs
9. Neutronic analysis of flux dispersion in a multi-layered, (D-T) driven hybrid blanket
10. Decrease of the LWR spent nuclear waste inventories and cumulative fissile fuel enrichment in hybrid reactor system
11. Hybrid reactor system
12. Fusion power from lunar resources
13. Apollo-advanced fuel fusion power reactor for the 21st century
14. Waste disposal assessment of HYLIEE-II structure
15. Mixed oxide fuel development
16. 2 system
17. Recent progress in fusion-fission (hybrid) reactor design studies
18. The homemade nuclear bomb syndrome
19. The fusion-fission fuel factory
20. Global energy scenarios and the potential role of fusion energy in 21st century
21. Realization of a flat fission power density in a hybrid blanket over long operation periods
22. Investigation of the neutronic potential of moderated and fast (D,T) hybrid blankets for rejuvenation of CANDU spent fuel
23. Rejuvenation of CANDU spent fuel in a hybrid blanket
24. Rejuvenation of LWR spent fuel in a catalyzed fusion-fission hybrid blanket
25. A numerical graphical power flattening method for source driven fast fission blankets
26. Decrease of the CANDU spent nuclear waste inventories in fusion-fission (hybrid) reactors
27. Neutronic parameters of a clindrical hybrid blanker driver by a simulated line source
28. Comparison of diffusion and transport theory for fast reactor shielding calculations
29. Potential of a catalyzed fusion driven hybrid reactor for the regeneration of CANDU spent fuel
30. 2 hybrid blanket driven by a (D,T) source
31. The effect of the spontaneous fission of plutonium-240 on energy release in a nuclear explosive
32. Status of research on Key LWR safety issues
33. Potential of a fusion-fission hybrid reactor using for various coolants to breed fissile fule for LWRs
34. 2) and different coolants for various volume fraction