Research Reactor Fuel Management 16 (RRFM 2012)

773 Pages·2016·37.92 MB·English

© 2012
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ISBN 978-92-95064-13-3

These transactions contain all contributions submitted by 16 March 2012.

The content of contributions published in this book reflects solely the opinions
of the authors concerned. The European Nuclear Society is not responsible
for details published and the accuracy of data presented.
Programme Committee

Edgar Koonen, SCK-CEN, Belgium (Chairman)
Gilles Bignan, CEA/RJH, France (IGORR Chair)
Pablo Adelfang, IAEA, Austria
Helmut Böck, TU-Vienna, Austria
Vladimir Broz, NRI Rez, Czech Republic
Philippe Auzière, AREVA NC, France
André Chabre, CEA, France
Heiko Gerstenberg, TU-München, Germany
Andrea Borio di Tigliole, LENA / Universita degli Studi de Pavia, Italy
Dominique Geslin, CERCA (AREVA Group), France
Istvan Vidovszky, AEKI, Hungary
José Marques, Insituto Tecnologico e Nuclear, Portugal
Gunter Damm, Jülich Research Center, Germany
Jocob de Vries, Delft Reactor Institute, The Netherlands
Stephen Curr, Rolls-Royce plc, United Kingdom
Peter Schreiner, GKSS Research Centre, Germany
Sandro Tozser, IAEA, Austria
Pascal Claude, AREVA, France (IGORR)
Douglas L. Selby, ORNL, United States (IGORR)
Jose Lolich, Balseiro Institute, Argentina (IGORR)
Danas Ridikas, IAEA, Austria (IGORR)
Monday 19 March 2012
2012 PROGRESS REPORT ON HEU MINIMIZATION ACTIVITIES IN
ARGENTINA

Pablo Cristini, Liliana de Lio, Daniel Gil, Alfredo G. Gonzalez,
Marisol López, Oscar Novara, Horacio Taboada

COMISIÓN NACIONAL DE ENERGÍA ATÓMICA
Av. Del Libertador 8250 (1429) Buenos Aires, Argentina

An extension of the original CNEA-NNSA DoE contract for the RA-6 reactor core
conversion was signed in March 2010 to enhance the final national HEU inventories
minimization. Previously, CNEA reserved a small inventory of HEU for further R&D uses in
fission chambers, neutronic probes and standards. This minimization comprises all fresh
and irradiated HEU remnant inventories coming from fuels and Mo99 targets fabrication
and irradiated HEU-oxides retained in production filters and solutions. Those inventories
are being recovered, down-blended into LEU and purified or, in few cases and due to cost-
benefit considerations, declared wastes.

CNEA has a R&D program focused on the development of the fabrication technology of U-
Mo monolithic (Zry-4 cladding) miniplates to support the qualification activities of the
RERTR program. Some monolithic 58% enrichment and LEU 8%Mo and U10%Mo
miniplates and plates were and are being delivered to INL-DoE to be irradiated in the ATR
reactor core. Full scale plates will take part of the ALT FUTURE irradiation at the BR II
Belgium reactor.

CNEA, a worldwide leader on LEU technology for fission radioisotope production is
providing Brazil with 1/3 of the national requirements on Mo99 by weekly deliveries. ANSTO
is firmly producing several fission radioisotopes batches by week. During November and
December 2011 the production in the new fission Mo-99 facilities of the Atomic Egyptian
Agency (AEA) in Inshas Atomic Center, Egypt was demonstrated. To support these
activities CNEA is refurbishing in Ezeiza Atomic Center a set of radiochemical cells where
the spent LEU based material retained in the filters of the Mo99 production facility of CNEA
along these last 10 years will be separed from wastes, recovered and purified to be
reutilized in this or in other nuclear applications. CNEA is strongly committed to improve the
diffusion of LEU target and radiochemical technology for radioisotope production and target
and their process optimization.

Future plans include:
o Development of waste separation techniques optimization.
o Fabrication and delivering to INL to be irradiated in the ATR core of complementary U-
8%Mo and U-10%Mo monolithic miniplates and development and fabrication of LEU very
high density monolithic and dispersed U-Mo fuel plates with Zr cladding for the FUTURE-
MONO experiment in the frame of the RERTR program.
o Optimization of LEU target and radiochemical techniques for radioisotope production and
waste separation, recovery and purification of irradiated LEU inventories contained in
fission Mo99 production filters..

1. Introduction:

In March 2010, a supplementary agreement between CNEA and NNSA-DoE to the original one -
involving the RA-6 reactor core conversion and the exportation to the US of 42 SNF in terms of
the SNF FRR Program- was signed by both parties. This was done in the frame of the efforts for
HEU minimization for civilian uses. From remaining HEU inventories, used in the past for fuel
and target fabrication, a small amount of it for R&D purposes was reserved by CNEA (for further
uses in fission chambers, neutronic probes and standards fabrication).

This minimization means the recovery, blending down and purification of fresh and irradiated
HEU inventories contained in scraps from fuel and target fabrication and in fission Mo99
production filters, or the disposition as waste of those few inventories whenever its recovery is
not advisable due to a cost-benefit consideration. These tasks are taking place and the
corrected deadline is December 2013.

2. New tasks on HEU minimization

Previously it was informed about the inventories classification into 6 groups (see Table 1)

Group U Mass 235 U Mass
Description Form Enrichment
Number (kg) (kg)
Irradiated Mo-99
Solid and
Targets And 1 1.928 89.73% 1.73
Liquid
Solutions
Gas / Solid
UF6 Cylinder 2 0.65 90.14% 0.59
(UO2F2)
Miscellaneous
Solids (alloys, 3 Solid 0.397 87.15% 0.346
metal)
Miscellaneous
4 Liquid 0.228 89.91% 0.205
Solutions
Materials declared
5 Solid 0.505 89.97% 0.453
waste to dispose
Ingot for MEU-
Mo/Zr Miniplates 6 Solid 0.344 88.66% 0.305
Fabrication
TOTAL 4.05 3.63

Tab. 1 HEU inventories

Regarding the progress of the tasks involved the present status can be seen in Table 2. Most of
them are already finished.
Group 1 comprises the refurbishment of a radiochemical laboratory (LFR lab), licensing of two
transport casks, for irradiated solutions and solids contained in cartridge filters, among of the
proper recovery, downblending and purification of the HEU inventory in the hot cells of the LFR
lab. This task is ongoing and the actualized deadline is December 2013.
Group 2 comprised the opening of a valve stucked 5A type cylinder containing a partial
hydrolized UF6 inventory, and the recovery, downblending and purification of the HEU inventory
into LEU. This task was finished on October 2010I
Group 3 implied the downblending of an HEU-Al inventory through cast melting. This task was
finished during September 2010.
Group 4 is made of several HEU solutions in acqueous and organic media. It was characterized
and partially downblended and partially wasted. It finished on May 2011.
Group 5: this inventory will be managed as waste and located in a new storage site, expecting to
finish this task by November 2012.
Group 6 included the blending down of a HEU ingot to 58% enrichment (at U235) to make ME
U-10%Mo and U-8%Mo based miniplates with Zr cladding. The agreement includes the
exportation to INL of miniplates for irradiation testing under the RERTR program. This task is
being accomplished with a first delivery made in April, 2010 and a final second one scheduled
for March 2012II. Finally, the remaining 58% enrichment inventory will be downblended to LEU-
10%Mo or LEU-8%Mo for full scale plate fabrication in the frame of the RERTR programme.

RECOVERED DECLARED
HEU MASS AND WASTE OR TO
GROUP DESCRIPTION
(Kg) DOWNBLENDED BE WASTED
(Kg) (Kg)
1 IRRADIATED MATERIAL 1,928 - -
5A CYLINDER
2 0,649 0,628 0,021
CONTAINING HEU-F6
3 SOLIDS 0,378 0,374 0,004
4 LIQUIDS 0,228 0,140 0,086
MATERIAL TO
5 0,505 0 0,505**
DECLARE WASTE
HEU FOR MEU-Mo
6 MINIPLATES & LEU-Mo 0,344 0,310* 0,034
PLATE FABRICATION
TOTAL 4,032 1.452 0.650

Tab. 2 Recovery and downblending into LEU

3. R&D on VHD fuels
The development of CNEA miniplates using monolithic UMo alloy cores involves the use of the
well known Zr alloy named Zircalloy-4 (Zry-4). Zry-4 is commonly used as nuclear fuels cladding
for nuclear power plants. In this case Zry-4 is employed as cladding in top, cover and frame. The
fuel material in the core is an alloy between Uranium (U) and Molybdenum (Mo) with different
contents of Mo (between 7-10% wt/wt). These Mo percentages are enough to retain the gamma
phase at low temperature but also not to penalize the reactor neutronics due to the capture
cross section of Mo95 isotope. The fabrication process includes hot and cold co-rolling.
Regarding hot co-rolling temperature between 575º - 660ºC the presence of some amount of
alpha plus delta phases were revealed by analyses. Also although hot rolling at 800-850ºC takes
place in gamma phase (with metastable bcc crystal structure), an important dog bone zoneIII and
Ux-Zyry-Moz phasesIV were found. To avoid both drawbacks an intermediate temperature at
700ºC was employed. Different rolling temperatures produced different effects:
 hot co-rolling at 650ºC shows a thin interlayer, but no difference in the composition of the
initial materials of cover, frame and fuel coupon.
 A small interlayer appear at the hot co-rolling at 700ºC which is a little different from the
650ºC, and a dog bone appeared at both coupon ends, but less relevant than the one
observed at 800ºC.
Rx, TEM, EDAX and microprobe analyses showed an interlayer between the Zry-4 covers and
the UMo coupon, formed with UxZry and UxZryMoz phases, that are in concordance with other
studiesV. Observed effects enabled improving the fabrication conditions.
Regarding the pack assembly it is essential to prevent gaps and oxygen presence as well as
impurities by achieving an intimate fit between Zry-4 and UMo fuel components and clean
surface contact. In other case that could be areas for potential debonds or undermine optimum
bonding during pack reduction.
In all cases TIG welding was employed in the same conditions, and different kind of interlayers
were observed. It can be concluded that this procedure do not contribute to the decomposition of
the UxMo.
Conclusions: for plate hot co-rolled at 800ºC a marked dog bone regions were observed at
each end of plate with are characteristic of the ductility difference between the U-xMo alloy and
Zry-4 at this temperature, but not roll speed effects were observed. This presence could be
diminished by using a longer lead or ends, but not disappeared. No dog bone zone was
observed in miniplates co-rolled at 650ºC.

LEU target and radiochemical technology for Mo99 and other fission radioisotopes
production:
It is by far the largest contribution of CNEA to HEU minimization for civilian uses. It was already
informed about the circumstances of the final cutoff to HEU supply for Mo99 production and how
CNEA found an adequate LEU replacement without changing its radiochemical technology.
CNEA has developed and is using high-density LEU-aluminum dispersion targets. The target
meat has a density of 2.9 gU/cm3 obtained by increasing the ratio of uranium aluminide to
aluminum in the target meat. The mass of U-235 in the target meat is about twice that of
conventional uranium-aluminum alloy targets.
CNEA was able to convert to LEU-based production in the same set of hot cells that were being
used for HEU-based production, without interrupting Mo-99 production. Targets are irradiated in
the RA-3 reactor at CNEA’s Ezeiza Atomic Center near Buenos Aires. Target processing is
carried out in a hot cell facility at the Ezeiza site. Process wastes are also managed at the site.
CNEA’s development showed that there are no technical barriers to conversion of fission Mo-99
production from HEU targets to LEU targets. Production using LEU targets is technically feasible
and is being carried out by CNEA in Argentina and by the Australian National Nuclear Science
and Technology Organisation. ANSTO is using CNEA’s radiochemical technology and LEU
targets to produce Mo-99.
This new LEU technology satisfies the most stringent requirements of quality for its use in
nuclear medicine applications. Mo-99 purity has been consistently higher than that produced
using HEU targets[VI]. Also in September 2005, CNEA began the regular production of high
quality fission I-131, a by-product of Mo-99 production, meeting also international quality
standards.
HEU-LEU production process comparison costs reveal that this new technology has no
significant overall cost of 10% [VII]. Due to the fact that CNEA was able to duplicate the LEU-
based radioisotope weekly production rate, since 2010 provides Mo99 to Brazil covering 1/3 of
the Brazilian market.
During November and December 2011 the production in the new fission Mo-99 facilities of the
Atomic Egyptian Agency (AEA) in Inshas Atomic Center, Egypt was demonstrated. This
achievement was the result of the collaboration between several technical groups of CNEA,
INVAP and AEA involved in this task, like experts in Mo99 production, product quality control,
design and facility building and radiochemical cells and laboratory equipment.
To support these activities CNEA is refurbishing a set of radiochemical cells where the spent
LEU based material retained in the filters of the Mo99 production facility along these last 10
years will be separed from waste, recovered and purified to be reutilized in this or in other
nuclear applications.
.
4. Conclusions:

FINAL HEU MINIMIZATION: CNEA is minimizing the remnants HEU inventories, both fresh and
irradiated from fuel and target fabrication scraps and fission RI production solutions and filters.
All these tasks are scheduled to finish during December 2012.
R&D ON LEU VHD FUELS: CNEA is actively supporting both R&D activities to achieve solutions
for core conversions.
LEU TECHNOLOGY FOR FISSION RI PRODUCTION: No technical, quality or financial reasons
make disadvantageous changing from HEU to LEU for fission Mo99 and other RI production.
CNEA leads LEU based isotope production technology, and with INVAP built all LEU-based
production systems in Australia and Egypt. This is by far the largest contribution of CNEA to the
HEU minimization for civilian uses.

I L. N. Aldave, H. Blanco Bello, A. A. Bonini, L. I. De Lio, L. A. Dell’Occhio, M. Falcón, T. Feijoo, A. Gauna, D. A. Gil,
A. Rodriguez y J. Valdez. 2010 RERTR International Meeting, Lisbon, Portugal, 10-14 October 2010.

II M.López, A. González, R. González, F. Rice, H. Taboada, D. Wachs. 2010 RERTR International Meeting, Lisbon,
Portugal, 10-14 October 2010.
III J. L. Snelgrove, G. L. Hofman, C. L. Trybus, and T. C. Wiencek. Development Of Very-High-Density Fuels By The Rertr Program.
19th International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Seoul, Republic of Korea. October 7-
10, 1996.
IV
M. K. Meyer, C. L. Trybus, G. L. Hofman, S. M. Frank, T. C. Wiencek. Selection and Microstructures of High Density Uranium
Alloys. 20th International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR) Jackson Hole, WY (US),
10 Oct 1997.

V J. L. Snelgrove, G. L. Hofman, C. L. Trybus, and T. C. Wiencek. Development Of Very-High-Density Fuels By The Rertr Program.
19th International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Seoul, Republic of Korea. October 7-
10, 1996
VI Durán,A. 2005. Radionuclide Purity of Fission Mo-99 Produced from LEU And HEU. A Comparative Study. 2005
International RERTR Meeting, Boston, Massachusetts, USA, November 6-10, 2005. Available at
http://www.rertr.anl.gov/RERTR27/PDF/S8-3_Duran.pdf.

VII Cestau D., A. Novello, P. Cristini, M. Bronca, R. Centurión, R. Bavaro, J. Cestau, E. Carranza. HEU and LEU cost
comparison in the production of molybdenum-99. 2008 International RERTR Meeting, Washington, DC, USA, 5-9
October 2008, and Cestau D., A. Novello, P. Cristini, M. Bronca, R. Centurión, R. Bavaro, J. Cestau, E.Carranza.
2007. HEU and LEU comparison in the production of molybdenum-99. 2007 International RERTR Meeting, Prague,
Czech Republic, Sep. 23-27, 2007. Available at http://www.rertr.anl.gov/RERTR29/PDF/6-4_Cestau.pdf
U.S. PROGRESS IN THE DEVELOPMENT OF VERY HIGH DENSITY
LOW ENRICHMENT RESEARCH REACTOR FUELS

M.K. MEYER, D. M. WACHS, J.-F. JUE, D.D. KEISER, J. GAN, F. RICE,
A. ROBINSON, N.E. WOOLSTENHULME, P. MEDVEDEV
Idaho National Laboratory
PO. Box 1625
Idaho Falls, ID 83415

G.L. HOFMAN, Y.-S. KIM
Argonne National Laboratory
9700 S. Cass Avenue
Argonne, IL 60439

ABSTRACT

The effort to develop low-enriched fuels for high power research reactors
began world-wide in 1996. Since that time, hundreds of fuel specimens
have been tested to investigate the operational limits of many variations of
U-Mo alloy dispersion and monolithic fuels. In the U.S., the fuel
development program has focused on the development of monolithic fuel,
and is currently transitioning from conducting research experiments to the
demonstration of large scale, prototypic element assemblies. These larger
scale, integral fuel performance demonstrations include the AFIP-7 test of
full-sized, curved plates configured as an element, the RERTR-FE
irradiation of hybrid fuel elements in the Advanced Test Reactor, reactor
specific Design Demonstration Experiments, and a multi-element Base
Fuel Demonstration. These tests are conducted alongside miniplate tests
designed to prove fuel stability over a wide range of operating conditions.
Along with irradiation testing, work on collecting data on fuel plate
mechanical integrity, thermal conductivity, fission product release, and
microstructural stability is underway.

1. INTRODUCTION

Analysis of the possibilities for very high enrichment fuel for converting the highest power
research reactors began in 1996. Investigation of possible fuel phases converged quickly on
uranium alloys stabilized in the gamma phase by molybdenum or a combination of niobium
and zirconium, or molybdenum with a ternary element addition as the primary candidates.
The RERTR-1 and RERTR-2 experiments were configured as a screening test of these
fuels, and began irradiation in the Advanced Test Reactor in 1997. [i] Results of these
experiments indicated superior performance of U-Mo fuel with more than 6 wt.%
molybdenum over U-Nb-Zr. Global interest in eliminating the use of high-enriched uranium
and the promising results from these experiments stimulated research around the world, and
aggressive programs began to develop and deploy this technology. Although U-Mo itself has
proven to be a robust fuel that is stable under irradiation to high burnup, challenges arose in
implementing this fuel in aluminum-based research reactor fuel system operating at high
power density. Unsatisfactory fuel plate swelling behavior in this system is caused by
irradiation stability of the fuel/matrix interaction layer. [ii,iii,iv] Progress has been made using
additives to the fuel matrix and coating of fuel particles to address this issue.

In the U.S., this problem and the requirement for very high uranium density led to the
development of monolithic fuel technology as an alternative to dispersion fuel. Monolithic
fuel is the primary candidate for U.S. LEU fuel development. As the U.S. program moves
from the research phase toward qualification of these fuels, there have been many