Core thermohydraulic design with 20% LEU fuel for upgraded research reactor JRR-3.

Sudo, Yukio; Ando, Hiroei; Ikawa, Hiromasa; Ohnishi, Naofumi

doi:10.3327/jnst.22.551

Cited by 8 publications

(4 citation statements)

References 4 publications

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“…The initial powers of 0.001W and 1.0W are the minimum power (source level power) and maximum power respectively at the start-up stage of KUR. Since the velocity difference in the low water velocity region has very low impacts on heat transfer and critical heat flux (Sudo et al, 1984), the initial average flow rate of the primary coolant is assumed to be 18 m 3 /h (5 cm/s in the core) at the natural circulation mode.…”

Section: Accidental Control Rod Withdrawal Transient At Natural Circulation Modementioning

confidence: 99%

Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

Shen

Nakajima

Unesaki

et al. 2013

Annals of Nuclear Energy

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The purpose of this study is to realize the full core conversion from the use of High Enriched Uranium (HEU) fuels to the use of Low Enriched Uranium (LEU) fuels in Kyoto University Research Reactor (KUR). Although the conversion of nuclear energy sources is required to keep the safety margins and reactor reliability based on KUR HEU core, the uranium density (3.2 gU/cm 3 ) and enrichment (20%) of LEU fuel (U 3 Si 2 -AL) are quite different from the uranium density (0.58 gU/cm 3 ) and enrichment (93%) of HEU fuel (U-Al), which may result in the changes of heat transfer response and neutronic characteristic in the core. So it is necessary to objectively re-assess the feasibility of LEU silicide fuel core in KUR by using various numerical simulation codes. This paper established a detailed simulation model for the LEU silicide core and provided the safety analyses for the reactivity insertion transients in the core by using EUREKA-2/RR code. Although the EUREKA-2/RR code is a proven and trusted code, its validity was further confirmed by the comparison with the predictions from another two thermal hydraulic codes, COOLOD-N2 and THYDE-W at steady state operation. The steady state simulation also verified the feasibility of KUR to be operated at rated thermal power of 5MW. In view of the core loading patterns, the operational conditions and characteristics of the reactor protection system in KUR, the accidental control rod withdrawal transients at natural circulation and forced circulation modes, the cold water injection induced reactivity insertion transient and the reactivity insertion transient due to removal of irradiation samples were conservatively analyzed and their transient characteristic parameters such as core power, fuel temperature, cladding temperature, primary coolant temperature and departure from nucleate boiling ratio (DNBR) due to the different ways and magnitudes of reactivity insertions were focused in this study. The analytical results indicate that the quick power excursions initiated by the reactivity insertion can be safely suppressed by the reactor protection system X. Shen et al. / Annals of Nuclear Energy 62 (2013) 195-207 2 of KUR in various initial power levels and different operational modes (natural circulation and forced circulation modes).No boiling and no burnout on fuel cladding surface and no blister in the fuel meat happens and KUR is safe in all of these reactivity insertion transients if the reactor protection system of KUR works in its minimum degree.

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Section: Accidental Control Rod Withdrawal Transient At Natural Circulation Modementioning

confidence: 99%

Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

Shen

Nakajima

Unesaki

et al. 2013

Annals of Nuclear Energy

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“…Therefore, a new core configuration with a neutron trap should be put forward. At present, most existing research reactors are designed with an inverse neutron trap, for example, JRR-3M [2,3] and CARR [4,5]. JRR-3M is a 20 MW research reactor and the thermal neutron flux reaches the peak, 2.3 × 10 14 n/cm −2 ⋅s, in the D 2 O reflector for the irradiation test.…”

Section: Introductionmentioning

confidence: 99%

“…Control rod with follower fuel element (6) Irradiation element (5) Be reflector CARR core JRR-3M core Figure 1: Cross-section view of CARR [4] and JRR-3M [2,3].…”

Section: Introductionmentioning

confidence: 99%

Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

Gong

Huang

Wang

et al. 2015

Science and Technology of Nuclear Installations

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Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

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“…These research reactors are characterized by large leakages due to their compact reactor design [3]. The Reactor Engineering Analysis Lab (REAL) at Tsinghua University has designed a research reactor with an inverse flux trapconverter target structure modeled after JRR-3M [4,5] and CARR (China Advanced Research Reactor) [6,7] to further investigate these research reactors.…”

Section: Introductionmentioning

confidence: 99%

Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

Huang

Gong

et al. 2017

Science and Technology of Nuclear Installations

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High thermal neutron fluxes are needed in some research reactors and for irradiation tests of materials. A High Flux Research Reactor (HFRR) with an inverse flux trap-converter target structure is being developed by the Reactor Engineering Analysis Lab (REAL) at Tsinghua University. This paper studies the safety of the HFRR core by full core flow and temperature calculations using the porous media approach. The thermal nonequilibrium model is used in the porous media energy equation to calculate coolant and fuel assembly temperatures separately. The calculation results show that the coolant temperature keeps increasing along the flow direction, while the fuel temperature increases first and decreases afterwards. As long as the inlet coolant mass flow rate is greater than 450 kg/s, the peak cladding temperatures in the fuel assemblies are lower than the local saturation temperatures and no boiling exists. The flow distribution in the core is homogeneous with a small flow rate variation less than 5% for different assemblies. A large recirculation zone is observed in the outlet region. Moreover, the porous media model is compared with the exact model and found to be much more efficient than a detailed simulation of all the core components.

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Core thermohydraulic design with 20% LEU fuel for upgraded research reactor JRR-3.

Cited by 8 publications

References 4 publications

Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

Reactivity insertion transient analysis for KUR low-enriched uranium silicide fuel core

Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

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