In-Pile Experiments in a Uranium-Zirconium … Experiments in a Uranium-Zirconium-Hydride TRIGA Fuel...

In-Pile Experiments in a Uranium-Zirconium-Hydride

TRIGA Fuel

In-pile Testing and Instrumentation for Development of Generation-IV Fuels and Materials

21-24 August 2012 Halden, Norway

Daniel Artur Pinheiro Palma, ScD dapalma@cnen.gov.br

(Researcher)

Amir Zacarias Mesquita , ScD amir@cdtn.br

(Researcher/ Senior Reactor Operator)

Brazilian Nuclear Energy Commission (CNEN)

Nuclear Technology Development Centre (CDTN)

Brazil

Guideline Abstract;

Motivation of this research;

IPR-R1 TRIGA Research Nuclear Reactor ;

Core coolant channels geometry and radial power distribution;

Heat transfer coefficient in the fuel gap;

Conclusion;

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• The heat generated by nuclear fission is transferred from fuel elements to the cooling system through the fuel-to- cladding gap and the cladding to coolant interfaces.

• The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were evaluated experimentally.

• A correlation for the gap conductance between the fuel and the cladding was also presented. As the reactor core power increases, the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling.

• Experimental results indicated that subcooled pool boiling occurs at the cladding surface in the reactor core central channels at power levels in excess of 60 kW.

Abstract

Motivation

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The objective of the thermal and hydrodynamic projects of the

reactors is to remove the heat safely, without producing excessive temperature in the fuel elements. The regions of the reactor core where boiling occurs at various power levels can be determined from the heat transfer coefficient data;

The thermal conductivity (k) of the metallic alloys is mainly a

function of temperature. In nuclear fuels, this relationship is more complicated because k also becomes a function of irradiation as a result of change in the chemical and physical composition. Many factors affect the fuel thermal conductivity. The major factors are temperature, porosity, oxygen to metal atom ratio, PuO2 content, pellet cracking, and burnup.

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The second largest resistance to heat conduction in the fuel rod is due to the gap. Several correlations exist to evaluate its value in power reactors fuels, which use mainly uranium oxide. The only reference found to TRIGA reactors fuel is General Atomic that recommends the use of three hypotheses for the heat transfer coefficient in the gap. The heat transfer coefficient (h) is a property not only of the system but also depends on the fluid properties.

The determination of h is a complex process that depends on the thermal conductivity, density, viscosity, velocity, dimensions and specific heat. All these parameters are temperature-dependent and change when heat is being transferred from the heated wall to the fluid.

An operational computer program and a data acquisition and signal processing system were developed as part of this research project to allow on line monitoring of the operational parameters

Motivation

IPR-R1 TRIGA Mark I

Open pool type reactor

November 11th, 1960

Power: 30 kW, 100 kW, 250 kW

4 % burnup

TRIGA

Training

Research

Isotopes

General Atomics

IPR-R1 TRIGA Research Nuclear Reactor

IPR-R1 TRIGA Research Nuclear Reactor

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Simplified core diagram of the IPR-R1 TRIGA

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Instrumentation for In-Pile Experiments

IPR-R1 TRIGA reactor cooling system diagram and instrumentation distribution

TRIGA stainless steel-clad fuel element

The Instrumented Fuel Element

Instrumented fuel element before insersion in the core

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IPR-R1 core top view with the instrumented fuel element in

ring B and one thermocouple inside the core channel

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Core temperature radial profile at 265 kW thermal power

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Fuel temperature as function of the reactor power in all core rings

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Core coolant channels geometry and radial power distribution

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Tcmq p

Nuclear Reactor Heat Transfer

Conduction

Fourier Equation

q = h A (Tsup – Tf)

Convection Newton Law

Pool boiling curve for water under atmospheric pressure

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Coolant properties and monophase convection coefficients – Single Phase

)TT(4

rqk

suro

2'''

g

Fourier equation

satsatsur TTT Wall superheat

259.0

sat )q(81.0T McAdams

Dittus-Boelter

4.0

p

8.0

w

w

spk

cGD

D

k023.0h

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Subcooled Nucleate Boiling Region

hsur = q” / ∆Tsat

Convection Newton Law

q = h A (Tsup – Tf)

• hsur convective heat-transfer coefficient ( fuel cladding to water);

• q" is the fuel surface heat flux ;

• ΔTsat is the surface superheat in contact with the water ;

Position

B1

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IPR-R1 TRIGA

Heat-transfer regimes

in the

fuel element surface

Subcooled Nucleate Boiling Region

hsur = q” / ∆Tsat

Overall fuel element thermal conductivity and cladding heat

transfer coefficient to the coolant

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HEAT TRANSFER COEFFICIENT IN THE FUEL GAP

Uncertainty ±7,5 %

Fuel element

configuration

Heat transfer

coefficient

through the gap

as a function

of the power

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Experimental fuel rod radial temperature profile in position B1 at

265 kW

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Position B1 at other

reactor powers

• Subcooled pool boiling occurs above approximately 60 kW on the cladding surface in the central channels of the IPR-R1 TRIGA core.

• However, the high heat transfer coefficient due to subcooled boiling causes the cladding temperature be quite uniform along most of the active fuel rod region and do not increase very much with the reactor power.

•The IPR-R1 TRIGA Reactor normally operates in the range from 100 kW until a maximum of 250 kW. On these power levels the heat transfer regime between the clad surface and the coolant is subcooled nucleate boiling in the hottest fuel element. Boiling heat transfer is usually the most efficient heat transfer pattern in nuclear reactors core.

• Another important aspect of the reactor operation safety is that it is far from the occurrence of critical heat flux. The results can be considered as typical for pool-type research reactor and are important for the possible upgrade of the IPR-R1 reactor to 500 kW, as well as to the project of the Brazilian Multipurpose Reactor (RMB in Portuguese). The RBM is being designed and will be a pool research nuclear reactor with about 20 MW of thermal power.

Conclusions

Thanks !

Brazilian Nuclear Energy Commission (CNEN)

Nuclear Technology Development Centre (CDTN)

Brazilian Council for Scientific and Technological Development (CNPq)

Research Support Foundation of the State of Minas Gerais (FAPEMIG).