Effect of cladding surface pre-oxidation on rod coolability under reactivity initiated accident conditions

Sugiyama, Tomoyuki; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 41(11), p.1083 - 1090, 2004/11

https://doi.org/10.3327/jnst.41.1083

The effect of cladding surface pre-oxidation on the rod coolability under reactivity initiated accidents was investigated. NSRR tests on irradiated fuel rods have shown higher rod coolability than that of fresh rods, which arose from suppressed DNB and early quench at the surface. To identify the dominant factor, possible factors such as pellet cracking and so on, were assessed. The most probable factor, the cladding pre-oxidation, was examined by pulse irradiation tests on fresh rods with three cladding surface conditions, no oxide layer, 1m and 10m-thick oxide layers. Temperature measurements showed increased thresholds for DNB and quench at the pre-oxidized surface, leading to a reduced film boiling duration. The shifts of the critical and minimum heat flux points could be caused by the surface wettability increase. In the present tests, the wettability change was probably dominated by the chemical potential change at the surface due to pre-oxidation. The test results indicate the effects do not depend on the oxide layer thickness, but on the presence of the oxide layer.

Feasibility study on high burnup fuel for Gas Turbine High Temperature Reactor (GTHTR300), 2

Katanishi, Shoji; Takei, Masanobu; Nakata, Tetsuo*; Kunitomi, Kazuhiko

Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.67 - 75, 2004/03

https://doi.org/10.3327/taesj2002.3.67

no abstracts in English

Experimental analyses results on the BFS 58-1-I1 critical assemblies

; Sato, Wakaei*; Iwai, Takehiko*

JNC TN9400 2000-096, 113 Pages, 2000/06

JNC-TN9400-2000-096.pdf:3.1MB

This report describes the updated analyses results on the BFS-58-1-I1 core. The experiment was conducted at BFS-2 of Russian Institute of Physics & Power Engineering (IPPE). The central region is "non-Uranium fuel zone", where only Pu can induce fission reaction. The non-U zone is surrounded by MOx fuel zone, which is surrounded by U0 fuel zone. Sodium is used for simulating the coolant material. As it was found that the lattice pitch had been incorrectly understood in the past analyses, all items have been re-calculated using the corrected number densities. Furthermore, significantly softened neutron spectrum in the central region caused problems in applying the plate-stretch model that has been established for fast reactor cores through JUPITER experimental analyses. Both keeping the pellet density and using SRAC library for the elastic cross section for lighter nuclides allow us to obtain reasonable analysis accuracy on the spectral indices that were measured at the center of the core. Application of such a cell model was justified through comparison among various cell models using continuous energy Monte-Carlo code MVP. It is confirmed that both the MOX zone and the U0 zone can be correctly evaluated by the plate-stretch model. Based on the updated cell calculation, both the effective multiplication factor (k-eff)and the spectral indexes agree well with the measured values. The transport and mesh-size correction is made for the k-eff evaluation. Those results also agree well within reasonable difference between those obtained by IPPE and CEA, which were obtained by using sub-group method or continuous-energy Monte Carlo code. Evaluation by the nuclear data library adjustment confirmed that the analyses results of the BFS-58-1-I1 core have no significant inconsistency with JUPITER experimental analyses results. Those results are quite important for starting BFS-62 cores, which will be analyzed in the framework of supporting program for Russian ...

lrradiation behavior and performance model of nitride fuel

; ;

JNC TN9400 2000-041, 29 Pages, 2000/03

JNC-TN9400-2000-041.pdf:1.18MB

Irradiation behavior and performance models were investigated in order to apply for nitride fuel options in feasibility study on fast breeder reactor and related recycle systems. (1)MechanicaI design of nitride fuel pin: The behaviors of fission gas release (increase of internal Pressure) and fuel-to-cladding chemical interaction (decrease of cladding thickness) are needed to evaluate cumulative damage fraction in case of fuel pin mechanical design. The behaviors of fission gas release and fuel-to-cladding chemical interaction were investigated from the past studies up to high burnuP, since the lower fission gas release in nitride fuel than in oxide fuel could contribute to reduce the plenum volume and result in the shortening of fuel Pin length. (2)Fuel pin smear density: The higher fuel smear density is preferred for the higher fissile density to improve the core characteristic. The behaviors of fuel pellet swelling were investigated from the past studies up to higher burnup, since the larger fuel pellet swelling in nitride fuel than in oxide fuel would restrict high burunp capability due to fuel-cladding mechanical interaction. (3)Compatibility of nitride fuel with high Temperature water: Compatibility of nitride fuel with high temperature water were investigated from the past studies to contribute water cooled fast breeder reactor options.

None

Saito, Hioraki*; Iriya, Yoshikazu*

JNC TJ8440 99-003, 156 Pages, 1999/03

JNC-TJ8440-99-003.pdf:2.72MB

no abstracts in English

Improvement of single-phase subchannel analysis code ASFRE-III; Verification analysis of fuel pin heat transfer model and pressure loss model

; Ohshima, Hiroyuki

PNC TN9410 97-104, 69 Pages, 1997/12

PNC-TN9410-97-104.pdf:1.56MB

As the part of the improvement of single-phase subchannel analysis code ASFRE-III, verification study about fuel-pin heat transfer model and flow resistance model of the code was carried out. Temperature distributions in a fuel pin predicted by the fuel-pin heat transfer model of ASFRE-III were compared with those calculated by the structure analysis code FINAS, which has been well validated and applied to various structure analyses, using the same boundary conditions. The comparison showed that the results by these two codes agreed with maximum difference of 1 %. and therefore the validity of the model was confirmed. With respect to the flow resistance model, distributed resistance model (DRM), which can enhance the consistent description of the fluid flow and wire-spacer interaction, was examined through analyses of two hydraulic simulation tests using the fifth mock-up fuel subassembly for the prototype LMFBR and the second mock-up fuel subassembly for the experimental rector. The calculated pressure difference between pressure measurement taps whose positions were near the top and the bottom of the fuel-pin bundle agreed with the measured data of both tests. The predicted pressure distribution in a horizontal cross section was also compared with the calculational result by the finite element analysis code SPIRAL and agreement was good.

Proceedings of the 25th anniversary meeting of the Alpha-Gamma facility

Kajitani, Yukio; ; Abe, Kazuyuki; Osaka, Masahiko; ; Hirosawa, Takashi; Koyama, Shinichi

PNC TN9440 97-004, 186 Pages, 1997/02

PNC-TN9440-97-004.pdf:21.19MB

The 25th anniversary meeting of the Alpha-Gamma Facility (AGF) at O-arai Engineering Center of PNC was held on February 7. The AGF started to examine irradiated materials on october 1 and fuel pins irradiated in the Dounreay Fast Reactor, DFR332/2 on December 1, 1971. The contents in this paper of the anniversary meeting are as follows. (1)25 years history and challenging plan for 2000 year. (2)Maintenance logbook of the facility, apparatus and manipulators for 25 years. (3)Recent results of melting temperature, thermal conductivity and lattice constants in irradiated MOX fuels. (4)Development on fission products release measuring apparatus and results of cold run tests. (5)Post irradiated examination results operated at the metallography cell in the Fuels Monitoring Facility (FMF). (6)Development on chemical analysis method for minor actinides (MA) in irradiated MOX fuels. (7)Refurbishment for MA containing MOX fuels, status and specifications for the fabrication and quality control apparatus.

None

Yoshida, Mamoru; ; Shobu, Nobuhiro; Aihara, Nagafumi;

PNC TN9700 95-001, 804 Pages, 1995/08

PNC-TN9700-95-001.pdf:30.97MB

no abstracts in English

Numerical prediction of turbulent heat transfer characteristics of a fuel rod with very small square ribs for high temperature gas-cooled reactors

Takase, Kazuyuki; Akino, Norio

The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.535 - 540, 1995/00

no abstracts in English

Measurements of the modified conversion ratio by gamma-ray spectrometry of fuel rods for water-moderated UO cores

Nakajima, Ken; ; Suzaki, Takenori

Nuclear Science and Engineering, 116, p.138 - 146, 1994/00

https://doi.org/10.13182/NSE94-A21489

no abstracts in English