atw 2018-09v3

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atw Vol. 63 (2018) | Issue 8/9 ı August/September

FUEL 444

uncoated zirconium experiences at

1,200 °C. At temperatures of 1,300 °C,

however, the Chromium- coated zirconium

alloy was stable for reasonable

lengths of time. Combined with the

lowering of zirconium oxidation at

normal operating temperatures, which

vastly reduced the formation of zirconium

hydrides, and therefore embrittlement,

the chromium- coated zirconium

provides significant performance

improvements during normal operation,

transients, design basis accidents

and beyond design basis accidents, as

compared to uncoated zirconium.

Similar tests were run with SiC at

temperatures from 1,600 °C up to

1,700 °C. These tests were terminated

only because of excessive corrosion of

the heater element. At 1,600 °C, the

SiC cladding was visually untouched.

At 1,700 °C, there were indications of

small beads on the surface, presumably

SiO 2 from the reaction of SiC

with steam, but on the whole, no

­significant deterioration of the SiC.

Changes are being made to the heating

rod to increase the flow of Helium

cover gas and to allow accurate weight

changes to be made on the SiC rodlets

so that kinetic data can be obtained.

3.3 Testing of Westinghouse

U 3 Si 2 ATF high-density fuel

U 3 Si 2 is a revolutionary material for

LWR fuel service because its inherent

thermal conductivity is much greater

than existing UO 2 -based fuel, resulting

in significantly lower pellet temperatures.

U 3 Si 2 -based fuel can also

have up to 17 percent greater uranium

density than UO 2 -based fuel, so considerably

more energy can be economically

realized from each individual

fuel assembly. However, due to

these differences, considerable data

is required on the behavior of U 3 Si 2

at LWR operating temperatures

(estimated to be from 600 °C and up to

1,200 °C during transients).

To obtain the necessary data,

U3Si2 fuel pellets were manufactured

at INL and put into rodlets in the ATR

in 2015. The first rodlets came out of

the ATR at the end of 2016 (Figure 3)

and post-irradiation examination

(PIE) was performed in the summer

of 2017 at INL [3]. The PIE results

indicate some small amount of

cracking that may have been due to

impurities within the U 3 Si 2 . Fission

gas release and swelling were both

essentially zero with an exit burnup

of 20 MWd/kgU. Considering the

ATR high heat generation rates (12 to

15 kW/ft), which are significantly

above the average of 5 kW/ft and peak

of 9 kW/ft normally found in LWRs,

this was exceptionally good behavior.

The next set of U 3 Si 2 pins is due out in

2018 and will have achieved a burnup

of 40 MWd/kgU.

U 3 Si 2 was tested for air and steam

oxidation and compared to UO 2 using

digital scanning calorimeters at both

the Westinghouse Fuel Fabrication

Facility in Columbia, South Carolina

(USA) [4] and at Los Alamos National

Laboratory (LANL) [5]. The Westinghouse

test results indicate that the

ignition temperatures for UO 2 and

U 3 Si 2 are between 400 °C and 450 °C.

The LANL results indicate an ignition

temperature of about 400 °C. The

reasons for this difference are being

studied. The heat and mass generated

by the oxidation of the U 3 Si 2 is considerably

higher than for UO 2 . The

effect of this difference in heat release

and mass on the stability of the rods

was investigated in rodlet tests in the

autoclaves in the Churchill facility

during the summer of 2017. Unacceptable

tube bulging was found and programs

are now underway to increase

the oxidation resistance of the U 3 Si 2 .

| | Fig. 3.

Neutron radiographs of 20 MWd/kgU U3Si2 pins from ATR. Note the lack of pellet cracking and

distortion.(Ref. 4).

It is noted, however, that ATF cladding

surfaces are much harder than zirconium

alloy cladding and grids, so it is

expected that the likelihood of grid to

rod fretting leakages will be greatly

reduced from the current ppm levels.

4 Accident scenario

evaluations

To assess and demonstrate the performance

of ATF materials in postulated

accident scenarios, Modular Accident

Analysis Program, Version 5 (MAAP5),

calculations were performed for chromium-coated

zirconium and SiC claddings

along with high-density fuels

for the station blackout scenario and

the Three Mile Island Unit 2 (TMI2)

small-break loss-of-coolant (LOCA)

scenario with replenishment of the

primary coolant [6].

The chromium-coated zirconium

option offers modest ATF gains

(~200 °C) before large-scale melting

of the core begins in beyond design basis

events, such as a long-term station

blackout. Though it would not prevent

the contamination of the PWR primary

loop due to ballooning and bursting at

about 800 °C to 900 °C, the chromiumcoated

zirconium option could prevent

a TMI-2 type of accident from extending

into the fuel meltdown phase and

prevent extensive contamination of

the containment and perhaps preserve

the nuclear plant. This is because,

although the Cr-coated Zr may begin

to fail as the temperature exceeds

1,400 °C due to eutectic formation, it

does not rapidly oxidize as uncoated

zirconium alloys do, and does not provide

a rapid energy input spike into the

core (Figure 4). Note that, in this case,

Iron-chromium-aluminum (FeCrAl)

was used to model the performance

of chromium-coated zirconium since

the temperature and oxidation performance

is about the same. The

results for the station blackout

scenario (Figure 5) indicate that

­fission ­products can be contained

within SiC cladding for up to two

hours longer than current Zr-based

cladding due to its higher temperature

capability (~2,000 °C decomposition

temperature). These two hours can

be used to implement additional

responses by the operators. The lower

pressure in the system due to minimal

hydrogen production (Figure 6.) increases

the chances that alternate

means to feed cooling water to the

core at about 40 gpm can result in

avoidance of fuel melting, indefinitely

extending the coping time as long as

the water flow continues. The SiC

cladding, of course, prevents any

Fuel

Westinghouse EnCore® Accident Tolerant Fuel ı Gilda Bocock, Robert Oelrich, and Sumit Ray

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