Postirradiation Examination, Analysis of Performance of HFIR Fuel Elements

89 Pages·1971·10.865 MB·English

ORNL-47 14
Contract No. W-7405-eng-26
METALS AND CERAMICS DIVISION
POSTIRRADIATION EXAMINATION AND EVALUATION OF THE
PERFORMANCE OF HFIR FUEL ELEMENTS
. A. E. Richt R. W. Knight G. M. Adamson, Jr.
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NOTICE
This report was prepared as
sponsored by the United States Government. Neither
the United States nor the United States Atomic Energy
Commission, nor any of their employees, nor any of
their contractors, subcontractors, or their employees,
makes any warranty, express or implied, or assumes any
legal liability or responsibility for the accuracy, com-
pleteness or usefulness of any information, apparatus,
product or process disclosed, or represents that its use
DECEMBER 97 1 would not infringe privately owned rights.
OAK RIDGE NATIONAL LABORATORY
Oak Ridge, Tennessee 37830
operated by
UNION CARBIDE CORPORATION
for the
U.S. ATOMIC ENERGY COMMISSION
DISCLAIMER

This report was prepared as an account of work sponsored by an
agency of the United States Government. Neither the United States
Government nor any agency Thereof, nor any of their employees,
makes any warranty, express or implied, or assumes any legal
liability or responsibility for the accuracy, completeness, or
usefulness of any information, apparatus, product, or process
disclosed, or represents that its use would not infringe privately
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DISCLAIMER

Portions of this document may be illegible in
electronic image products. Images are produced
from the best available original document.
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CONTEJTCS
Page
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. . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Abstract 1
. . . . . . . . . . . . . . . . . . . . . . . . . . .
Introduction 1
General Description of the HFIR Fuel Elements and Reactor
. . . . . . . . . . . . . . . . . . . . . .
Operating Conditions 3
Selection and Operational History of the Examined HFIR
. . . . . . . . . . . . . . . . . . . . . . . . . .
Fuel Elements 12
. . . . . . . . . . . . . . . . . . .
Postirradiation Examination 15
. . . . . . . . . . . . . . . . . . . . .
General Objectives 15
. . . . . . . . . . . . . . . . . . . . . . . . . .
Procedure 16
. . . . . . . . . . . . . . . . . . . . . . . . . . .
Results 17
. . . . . . . .
Visual Inspection of the Fuel Assemblies 17
. . . . . . . .
Element Length and Diameter Measurements 21
. . . . . . . . . . . .
Fuel Plate Spacing Measurements 21
. . . . . . . . . . . . . . . . . . .
Fuel Plate Removal 23
. . . . . . . . . .
Visual Inspection of the Fuel Plates 25
. . . . . . . . . . . . . . . . .
Gamma Scanning Results 32
. . . . . . . . . . .
Fuel Plate Thickness Measurements 34
Fuel-Plate Sectioning and Analytical
. . . . . . . . . . . . . . . . .
Burnup Determinations 39
. . . . . . . . . . . . . . .
Metallographic Examination 43
. . . . . . . . .
Pr e irradiat ion Blister- Annealing Tests 55
. . . . . . . . . . . . . . . . . . . . . . . . . . . .
Discussion 59
. . . . . . . . . . . . . . . . . . . .
Dimensional Stability 59
. . . . . . . . . . . . . . . . . . . .
Structural Integrity 61
. . . . . . . . . . . . . . . . . . . . . . . . . .
Corrosion 61
. . . . . . . . . . . . . . . . . . . . .
Additional Comments 63
. . . . . . . . . . . . . . . . . . . . . . . . . . .
Conclusions 63
. . . . . . . . . . . . . . . . . . . . . . . . .
Acknowledgments 64
. . . . . . . . . . . . . . . . . . . . . . . . . . . .
AppendixA 67
. . . . . . . . . . . . . . . . . . . . . . . . . . . .
Appendix13 74
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AppendixC 77
. . . . . . . . . . . . . . . . . . . . . . . . . . . .
AppendixD 82
. . . . . . . . . . . . . . . . . . . . . . . . . . . .
AppendixE 84
POSTIRRADIATION EXAMINATION AND EVALUATION OF THE
PERFORmNCE OF HFIR FUEL ELEMENTS
A. E. Richt R. W. Knight G. M. Adamson, Jr.
ABSTRACT
This report describes the results of the postirradiation
examination and evaluation of the prior performance of four
spent fuel elements fromthe High Flux Isotope Reactor. Two ele-
ments had been immersed in boiling deionized water for 24 hr
before being operated for 2046 MWd. The other two were not pre-
treated and operated for slightly over 2300 MWd (the normal
reactivity lifetime). These elements were examined to determine
their overall condition and detect any unexpected problems that
might compromise continued operation of the present fuel ele-
ments to their full reactivity lifetime.
A l l four fuel elements were in excellent condition. Gamma
scanning and analytical burnup determinations showed the time-
integrated power and burnup distribution within these elements
to agree very well with design predictions. No significant
changes in the dimensions of either the fuel assemblies or the
individual fuel plates were indicated. The corrosion-product
film on the surfaces of the fuel plates was somewhat thicker
than expected, particularly on the pretreated fuel elements.
r
Consequently, this pretreatment of the fuel elements is undesir-
able. Even with the thicker film, however, the elements appeared
to perform quite satisfactorily; extensive visual and metallo-
graphic examination of fuel plates from these elements showed
no indications of any type of structural damage that could be
considered indicative of an actual or incipient failure of the
fuel elements.
Since all four elements were in such good condition, we
conclude that the current HFIR fuel elements w i l l perform quite
satisfactorily throughout their normal lifetime under present
operating conditions. The useful lifetime is apparently limited
only by nuclear reactivity and not by any irradiation damage.
INTRODUCTION
The High Flux Isotope Reactor (HFIR) currently operating at Oak
Ridge National Laboratory was designed and constructed essentially for
a single purpose - to produce annually milligram quantities of
2
various transplutonium isotopes from a rather limited supply of 242Pu
feed material. Preliminary analysis of the transplutonium production
schemes1 and studies of possible reactor ty-pes2J3 indicated that the
use of a highly enriched, beryllium-reflected, light-water cooled and
moderated, flux-trap reactor having a maximum unperturbed thermal-
loT5
neutron flux of about 5 X neutrons sec-l would be an economical
and technologically feasible way of achieving the desired production
rates. The reactor power required to obtain such a high flux was
about 100 MW, which appeared to be consistent with the capital f'unds
available. As the reactor concept was developed, a nominal power level
of 100 MW became one of the design criteria; however, since the trans-
plutonium production rates were found to be quite sensitive to the
neutron flux level, considerable emphasis was placed upon achieving the
highest practical flux level within the flux trap. As a result of
these efforts, the present design4 of the HFIR represents a significant
advancement in increasing the performance capabilities of research and
test reactors. For example, at the design power level of 100 MW, the
peak thermal and nonthermal neutron fluxes in the fuel region of the
HFIR are 2 and 4 X lo1* neutrons sec'l, respectively; the maximum
power density is 4.26 MW/liter; and the peak heat flux is about 500 W/cm2
(1.6 X lo6 Btu hr-' ft-2).
Thin, aluminum-base fuel plates for the HFIR fuel elements were
selected because minimum development, operating, and capital investment
costs were expected. However, HFIR fuel plates would be subjected to
heat fluxes, temperatures, and burnup levels considerably higher than
those of the more conventional types of research reactors. Moreover,
'J. A. Lane et al., High Flux Isotope Reactor Preliminary Design
.
Study, ORNL-CF-59-2-65 (March 1959)
2R. D. Cheverton, HFIR Preliminary Physics Report, ORNE3006
(October 1960).
3N. Hilvety and T. G. Chapman, "Thermal Design of the HFIR Fuel
Elements," pp. 138-151 in Research Reactor Fuel Element Conference,
September 17-19, 1962, Gatlinburg, Tennessee, TID-7642, Book 1.
.
'2'. T. Binford and E. N. Cramer, The High Flux Isotope Reactor,
Volume I. Functional Description, ORNL-3572 (May 1964).
.
3
in overall design, the HFIR fuel assemblies represented a rather radical
departure from those of the other reactor systems.
As of August 1971, over 60 fuel cores have operated at the full
design power level of 100 MW to the end of nuclear life without any
significant operational problems and with no indications of a f'uel ele-
ment or fuel plate failure. While such an outstanding performance
record attests to the satisfactory design of the HFIR fuel elements, we
felt that some of the spent fuel assemblies should be subjected to a
detailed postoperational examination to more thoroughly evaluate the
performance of these elements. The results of the examination of these
fuel assemblies are the subject of this report.
GENERAL DESCRIPTION OF THE HFIR FUEL ELEMENTS
AND REACTOR OPERATING CONDITIONS
The design and performance characteristics of the HFIR fuel elements
differ considerably from those of the more conventional research reac-
tors. The reactor core (Fig. 1) basically consists of four concentric
cylindrical regions arranged from inside to out in the following order:
1. the central flux-trap region, which is about 5 in. in diameter,
normally contains an assembly of target rods and which initially
contained a plutonium dispersion;
2. the fuel region, composed of two concentric, annular fuel
as s emblies ;
3. the control region, consisting of two concentric, cylindrical
control elements each about 1/4 in. thick;
4. the beryllium reflector, which is about 43 in. OD X 12 in. thick.
During operation the coolant water at 600 psi flows downward
through the fuel elements at about 55 ft/sec (13,500 gpm). Under these
conditions the pressure drop through the elements is about 105 psi.
The coolant inlet temperature is maintained at 110 to l.2O0F, while the
outlet temperature is 150 to 160°F. The pH ofthe coolant water is
maintained at 5.0 by nitric acid additions to minimize corrosion of the
aluminum fuel plates. At the design power level of 100 MW, the average
and maximum heat fluxes in the fuel assemblies are calculated to be about
ORNL- LR- DWG- 65776A .
.
Fig. 1. Schematic Representation of the HFIR Core Components.
5
lo6
0.8 and 1.6 X Btu hr-’ ftm2,r espectively, while the average and
maximum power densities are calculated to be 1.92 and 4.26 MW/liter,
.
respectively
The reactor operating experience has shown that the fuel elements
.
have a useful lifetime of about 23 days ( 2300 MWd)
The overall design characteristics of the HFIR fuel elements are
shown in greater detail in Fig. 2. The inner or outer fuel element
consists of an annular array of 0.050-in.-thick aluminum-base fuel
plates, which are attached to grooved cylindrical side plates by circum-
ferential TIG weld beads. The fuel plates in both elements are pre-
formed to an involute curvature to maintain a constant 0.050-in. coolant
channel spacing between the plates. The fuel plates of both fuel ele-
ments are of a sandwich-type construction, consisting of a fuel core
composed of a dispersion of U3O8 in aluminum, which is clad on all
surfaces with aluminum alloy 6061. A unique characteristic of the HFIR
fuel plates is the incorporation of a nonlinear fuel gradient across the
fuel plate. As shown in Fig. 3, this is achieved by varying the thickness
of the fuel core. The purpose of this fuel gradient is apparent when
one considers the radial neutron flux distribution in the HFIR. As
shown in Fig. 4, the thermal neutron flux peaks very sharply near the
inner and outer edges of both fuel elements. If the fuel was uniformly
distributed in a radial direction, the power distribution would also
peak sharply in these areas. However, if the fuel distribution is varied
inversely with the flux profile, the power distribution across the fuel
plates can be essentially flattened. Since this reduces the peak-to-
average power generation ratio, the fuel plates can then be allowed to
operate at a higher average power density. The rather unique control
rod system used in HFIR also helps to reduce the peak-to-average power
ratio. In normal operation the inner control cylinder moves downward
while the outer cylinder simultaneously moves upward to increase reac-
tivity. This maintains axial symmetry of the power density and reduces
the power peaking normally encountered at the upper and lower ends of
the fuel core.
6
k
0 .
.
.067" dia
.O5O" Thick Plates
,050" Thick Coolant Gaps
Fig. 2. Sketch of the HFIR Fuel Elements Showing Nominal Dimensions
and Overall Configuration.