Closure to “Discussion of ‘A Mathematical Model for Transient Subchannel Analysis of Rod-Bundle Nuclear Fuel Elements’” (1974, ASME J. Heat Transfer, 96, p. 565)

This paper presents a mathematical method for analyzing transient flow and enthalpy transport in rod-bundle nuclear fuel elements during both boiling and nonboiling conditions. A mathematical model is formulated by dividing the bundle flow area into flow subchannels that are assumed to contain one-dimensional flow and are coupled to each other by turbulent and diversion crossflow mixing. The mathematical model neglects sonic velocity propagation and neglects temporal and spatial acceleration in the transverse momentum equation. A semiexplicit finite-difference scheme is used to perform a boundary-value solution where the boundary conditions are the inlet enthalpy, inlet flow rate, and exit pressure. Calculations are presented to show the effect of rapid changes in heat flux, inlet enthalpy, and inlet flow rate on the subchannel flow and enthalpy distribution in rod bundles.

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COBRA-II: A DIGITAL COMPUTER PROGRAM FOR THERMAL-HYDRAULIC SUBCHANNEL ANALYSIS OF ROD BUNDLE NUCLEAR FUEL ELEMENTS.

10.2172/4183936 ◽

1970 ◽

Cited By ~ 11

Author(s):

D S Rowe

Keyword(s):

Computer Program ◽

Nuclear Fuel ◽

Digital Computer ◽

Rod Bundle ◽

Fuel Elements ◽

Subchannel Analysis

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Design Of Dry Cask Storage For Serpong Multipurpose Reactor Spent Nuclear Fuel

Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir ◽

10.17146/urania.2018.24.1.3694 ◽

2018 ◽

Vol 24 (1) ◽

Author(s):

Dyah Sulistyani Rahayu ◽

Yuli Purwanto ◽

Zainus Salimin

Keyword(s):

Heat Transfer ◽

Nuclear Fuel ◽

Spent Nuclear Fuel ◽

Construction Material ◽

Trial And Error ◽

Spent Fuel ◽

Concrete Wall ◽

Transfer Resistance ◽

Steel Liner ◽

Fuel Elements

DESIGN OF DRY CASK STORAGE FOR SERPONG MULTI PURPOSE REACTOR SPENT NUCLEAR FUEL. The spent nuclear fuel (SNF) from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF). At present there are a remaining of 245 elements of SNF on the ISSF,198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decayat a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg2. The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectedly. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method basedon SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister. The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 oC and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i.e. rack canister no 3, 8 and 10. The value of I0 from the rack no 3, 8 and 10 are 434.307; 446.344; and 442.375 mrem/h respectively. The total heat decay from rack canister no 3,8 and 10 are 179.640 ; 335.2; and 298.551 watts. The result of the trial and error calculation indicates that the rack canister no 3, 8 and 10 need the thickness of concrete shielding of 0.1912, 0.1954 and 0.1940 m respectively.Keywords: heat and radiation decay, spent fuel , storage cask.

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Large-Scale Pulsation Detection by Means of Temperature Measurements

Journal of Heat Transfer ◽

10.1115/1.2969262 ◽

2008 ◽

Vol 130 (11) ◽

Cited By ~ 2

Author(s):

N. Silin ◽

V. Masson ◽

A. Rauschert

Keyword(s):

Nuclear Fuel ◽

Large Scale ◽

Axial Flow ◽

Temperature Measurements ◽

Time Resolved ◽

Structure Information ◽

Rod Bundle ◽

Nuclear Fuel Assembly ◽

Large Scale Flow ◽

Fuel Elements

In the present work we explore the potential of time-resolved temperature measurements to obtain information on large-scale pulsations in a rod bundle geometry with axial flow. Large-scale flow pulsation is the phenomenon that dominates the turbulent mixing between the subchannels of rod bundles, which explains why it is of great importance for the design or assessment of nuclear fuel elements. The objective of the present work is to determine the characteristics of large-scale pulsations that can be used for the verification or validation of computational fluid dynamics code results. The method proposed is to generate a temperature gradient across the location of flow pulsations and to measure the time-varying temperature field downstream. Pulsation characteristic times, lengths, and traveling speed have been obtained. This study has been performed in a rod bundle similar to a nuclear fuel assembly and the results obtained are in good agreement with previous works on similar geometries. The technique can be applied to obtain additional large-scale structure information in test sections designed for thermal measurements, in situations where convection is dominated by these structures.

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