M. R. FINLAY and M. I. RIPLEY

Materials and Engineering Science  Australian Nuclear Science and Technology Organisation (ANSTO)  PMB 1 Menai, NSW 2234 Australia

Abstract

In the two years since the last ANA conference there have been marked changes in the  research reactor fuel scene. A new low-enriched uranium (LEU) fuel, “monolithic” uranium  molybdenum, has shown such promise in initial trials that it may be suitable to meet the  objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to  converting all US and Russian research reactors to LEU by 2012. Development of more  conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is  entering the final qualification stage of multiple full sized element irradiations. Despite this  progress, the original 2005 timetable for UMo fuel qualification has slipped and research

reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The  operators of the Swedish R2 reactor have been forced to pursue the direct route of  qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal  requirements of the Swedish law. The LTA has recently been fabricated and is expected to  be loaded shortly into the R2 reactor. W e  present an update of our previous ANA paper [l]

and details of the qualification process for UMo fuel.

1. INTRODUCTION

There has been a worldwide movement, largely  driven by the US Reduced Enrichment for Research  and Test Reactors (RERTR) program that was  introduced in 1978, to reduce the enrichment of  uranium used in research reactors to less than 20%

U. However, a reduction in uranium enrichment  requires a concomitant increase in uranium density  to maintain an equivalent neutron flux. Uranium  density has been increased in response to the move  to low enriched uranium (LEU) by improvements in  fabrication technology and the development of new  dispersion type fuels. Dispersion type fuels contain

a homogeneous distribution of fuel particles in an  inert matrix, which has generally been aluminium,  because of its attractive thermal, neutronic,

mechanical and corrosion properties. In 1988, the  US Nuclear Regulatory Commission approved the  use of a dispersion of LEU uranium silicide (U3Si2)  in an aluminium matrix up to a density of

4.8 gU/cm3. ANSTO’s Replacement Rescarch  Reactor (RRR) will commence operation using this  fuel.

Some research reactors however, require a higher  uranium density of 8-9 gU/cm3 to enable them to

23s

convert to LEU without an unacceptable reduction in  neutron flux. Given that fabrication techniques for  uranium silicide were almost fully optimised, the  only option was to pursue a different fuel with the

potential to Qffer higher uranium density. A number  of fuels were identified and evaluated by the RERTR  program. The only fuel to show serious promise is a  high density uranium molybdenum alloy.

A further restriction on research reactor operators  outside the US that needed to convert to LEU has  been the absence of a facility willing to reprocess or  otherwise dispose of their spent silicide fuel beyond  a strictly limited quantity. Many research reactor  operators have taken advantage of a US Department

of Energy (DOE) program, the Foreign Research  Reactor Spent Nuclear Fuel Acceptance (FRRSNFA)  Program, to take back and dispose of their spent fuel.  The scheduled limitation of this program to receive

only fuel irradiated prior to May 2006 will remove  this convenient disposal option. For these reactors to  continue operation, the alternatives are either to  dispose of their fuel in

a country other than the US,

or convert to UMo fuel. UMo fuel is suitable for  reprocessing [2] and will be accepted by at least one  reprocessor (COGEMA). Therefore, the date of  May 2006 has been seen by some reactor operators

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as a deadline for the qualification of the uranium  molybdenum fuel.

2. CURRENT FUELS

HIFAR and most research reactors of its generation  have fuel assemblies that consist of fuel plates  containing highly enriched uranium (HEU). HEU  fuels are advantageous because the increased

percentage of the fissile isotope 235U permits more  compact cores and hence gives the high neutron  fluxes required for research reactor applications. In  addition, with increasing enrichment a concomitant  reduction in 238U occurs: 238U not only contributes  very little in the fission process, but also acts as a  neutron absorber.

In conventional HEU fuels, the fuel meat is an alloy  of uranium and aluminium, known as uranium  aluminide (UA1,-AI). Pure uranium metal is not  used because it exhibits an unacceptably high rate of  growth and swelling under irradiation. Aluminium is  an ideal alloying element because of its low neutron  capture cross-section, high thermal conductivity,  workability and high corrosion resistance in water.

The requirement to shift from HEU to LEU

presented research reactor operators with problems  for the continued use of UA1,-AI fuel. The main  limitation was the low density of uranium in the  UA1,-A1 fuel meat, -l.7gU/cm3,

as shown in

Table 1. An increased uranium density is required in  LEU fuels to increase the number of fissile 235U  atoms to compensate for the deleterious effects of  the 238U atoms. The uranium density in the fuel meat

is lower than the phase density because the fuel meat  contains only 10-50% of the phase by volume, with  the balance being aluminium. The practical limit for  aluminide fuel was found to be -2.4 gU/cm3. In

order to convert existing research reactors

successfully to LEU without unacceptable penalties  in neutron flux, safety or economics, it was necessary  to develop higher density uranium alloys.

Uranium silicide alloys with a density of initially 3.8,  and then 4.8, gmU/cm3 were developed by the  RERTR program and qualified in 1988 [ 5 ] .  These  fuels are now the norm for high power research  reactors. Most of these reactors use conventional  plate type fuel with fuel meat of U3Si2. U3Si is less

suitable as a fuel than U3Si2, as under some

conditions it can exhibit breakaway swelling.  Nonetheless, its higher density is attractive and it is  used in pin-type fuels in the Canadian Maple design  reactors. Details of the metallurgy and fabrication  procedures for these alloys have been given  previously [ 11.

3. URANIUM MOLYBDENUM (UMO) FUEL

The limit on the maximum achievable uranium  density in U3Si2 triggered a search for other uranium

compounds that could offer even higher uranium  density. A number of other compounds were  evaluated including u6FC, U6Mn and additions of

a

third element to UMo to form ternary compounds.  Although the ternary additions did provide slight  improvements, the benefit was not sufficient to  warrant further effort. In 1996, the RERTR program

selected UMo as the primary candidate to be further  evaluated in a thorough campaign to demonstrate its  irradiation performance as a dispersion fuel. The  properties of stable fuel compounds are shown in  Table 1 .   In the early stages of the campaign, the RERTR  worked closely with the Korean Atomic Energy  Research Institute (KAERI) which had developed  and refined a technique to fabricate UMo

powder [4]. UMo is not a brittle intermetallic like  U3Si2 but, rather, is a ductile alloy. As a result, it is  difficult to produce an acceptable yield of particle  sizes in the desired range by mechanical means. The  KAERI process forms spherical particles by virtue of  a stream of a liquid melt of UMo hitting a spinning  disc, from which droplets spin off and solidify in a

vacuum. Work to characterise the powder and the  performance of plates fabricated with the powder  was adversely aÃ¢ÂÂ¬fected by a disputed patent gained  by KAERI in the US and other countries. The patent

not only covered the atomisation process but any  dispersion fuel containing spherical powder. The  patent has been largely responsible for the 2 year  delay currently experienced by the program.

The restrictions imposed by the patent and the  subsequent delays spawned efforts to produce UMo  powder by alternative means. The most promising  development has been achieved by the Argentine  Nuclear Comisi6n Nacional de Energia Atcimica

(CNEA). The process is referred to  as hydride-

milling-dehydride (HMD) as a result of the

discovery that UMo will absorb large amounts of  hydrogen in the gamma phase to form brittle  hydrides at low temperatures and pressures [6].  Such is the promise of this fuel powder that it will be  trialled in the full-sized element irradiations to be  conducted in the HFR Petten (see Table 2)

4. QUALIFICATION OF UMo FUEL

The qualification of UMo fuel is an international  effort carried out under programs in the United  States (RERTR program - Argonne National  Laboratory), France - French UMo Group (CEA,  CERCA, COGEMA, Framatome-ANP, and  Technicatome), and Argentina (CNEA).

A fuel is generally regarded as ‘qualified’ once the  information that is necessary to approve its use by a  regulatory authority has been documented in a  qualification report. The licensing stage is the more

specific process where an operator of a reactor

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applies to the regulator to use the fuel under defined  conditions. The US NRC previously reviewed a  generic report on U$i2 and issued an approval -  NUREG 1313 [5] - for use in order to facilitate the

licensing of a number of reactors using the same fuel  within a window of defined operating conditions.  Each operator must still demonstrate to its regulator  that use of the fuel is within the defined conditions.

A similar approach to the qualification of U3Si2 will  be followed for UMO. The RERTR program will  submit a qualification report to the US NRC on UMo  fuel that will include contributions from each  member of the program. The process includes:

Development of industrial processes to produce  UMo particles and fuel plates.

Testing in-reactor and ex-reactor to provide basic  data on miniature and full-sized plates.

Assessing reprocessibility.

Developing an understanding of fuel behaviour  under operational conditions, including the use of  models to predict fuel behaviour.

Fabrication and irradiation  of full-sized fuel

elements. This test provides good statistics  because of the irradiation of many plates

fabricated under industrial scale conditions. It  also serves as a useful qualification process of  fabricators. Typically, one element is irradiated  to >70% average burn-up to provide assurance  that a margin of safety exists for the operators,  who usually limit burn-up to -55%.

The RERTR program has irradiated more than 120  mini-plates to assess the effect of molybdenum  composition, different powder production methods,  different fission rates, and different temperatures.  The French program has irradiated 13 full sized  plates in three different reactors with similar goals.

Table 3 summarises the five different irradiation  experiments conducted by the RERTR program.  The RERTR 1 and RERTR 2 experiments of low  uranium density were largely the same, the

difference being -40% burn-up and -70% burn-up  respectively. The RERTR 3 experiment was  performed with higher uranium density and higher  temperatures. The RERTR 4 and RERTR 5

experiments aimed to fully evaluate a range of  loadings and temperatures more likely to be  experienced in normal research reactor operation and  were taken to -80% and -50% average burn-up

respectively.

A less extensive, parallel program is being

conducted by the French using full sized plates. The  details of the program are presented in Table 4. It is  clear that the French program is more aggressive and  targets qualification of fuel with a density of

8 gU/cm3. This effort is largely driven by the need  for a high density fuel for the proposed Jules

Horowitz Reactor.

The failure of the 35% enriched

plates from the UMUS experiment however, means  that fuel performance may be limited by oxide  thickness on UMo plates. The attractiveness of the  monolithic fuel described below may replace the  need to pursue the qualification of dispersion fuel at  8 gu/cm3.

The final stage of the qualification tests, which  consists of full-size fuel element irradiations, is  scheduled to commence in early 2004. The details  and schedule of each test are shown in Table 2 and  Figure 1 respectively. As the tests are proceeding,

work will commence on a preliminary qualification  report with all the necessary information except the  results of the full-sized element irradiations that are  planned to be available by the end of 2005. The US

RERTR program has secured in-principle agreement  from the US NRC to issue generic approval for UMo  in the same manner as for U3Si2 (NUREG 1313).  The NRC will be asked to begin by reviewing the  preliminary report since it will contain almost all the  necessary information., It will take approximately  one year to complete the full review and issue a  license.

In Sweden, the law requires the operator of a nuclear  reactor to have a demonstrated spent rue1 disposal  route in place in order to issue an operating license.  The fuel from the R2 reactor is currently returned to

the US under the FRRSNFA (takeback) program.  The projected delay in the qualification of UMo  beyond the May 2006 date could cause the reactor to  shut down. The operators of R2 have therefore been

forced to pursue a more direct path to qualification  and licensing of UMo fuel to ensure that it is

available for their reactor beyond this date. They  plan to irradiate a UMo lead test assembly (LTA)  and submit the#, results in conjunction with existing  knowledge about UMo fuel to their regulator for

a

licence. The LTA was fabricated by CERCA and  was ready to be loaded into the reactor in March of  this year, but regulatory approval has not yet been  issued (as of September 2003).

5. “MONOLITHIC” UMO FUEL

Monolithic UMo fuel uses a thin foil of U-10Mo  instead of a dispersion of UMo fuel particles in a  matrix of aluminium. The idea that led to trials of a  thin foil came from knowledge that a significant  reaction occurs between each UMo fuel particle and

the surrounding aluminium matrix (see Figure 2).  While the phase is stable, with relatively low fission-  induced swelling, it does occupy a larger volume  than its UMo and aluminium constituents and

therefore contributes to overall swelling. The  formation of the interaction layer is temperature  dependent. Under conditions of high temperature or  high burn-up, the formation of the interaction layer

in dispersion fuel is sufficiently extensive to

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completely deplete the fuel meat of aluminium. In  contrast, the use of a thin foil minimises the surface  area and resultant swelling. The initial trials in the  RERTR 4 experiment were very promising [3].  In addition to a reduction in swelling, there are a

number of other potential benefits, including  expected reduction in fabrication costs. The most  appealing benefit arises from the removal of the  aluminium matrix, which permits much higher  uranium loadings in the same volume. Assuming  fabrication difficulties can be overcome; it may be  possible to achieve uranium densities in the fuel of

-16 gU/cm3. Such  a development is of great

significance when considered in terms of the Joint  Declaration signed by Presidents Bush and Putin at  the May 2002 summit held in Moscow. The

declaration states that the two countries will  accelerate plans to reduce the use of weapons grade  fissile material declared to be in excess of defence  needs. In that declaration was a commitment by the

US to convert the five largest HEU-fuelled research  reactors used in civilian programs. These five  reactors could not be converted to LEU using any of  the dispersion fuels either currently available or

expected to be available in the coming few years.

The head of the US RERTR program, Travelli,  recently stated [8] that the development and  qualification program of such a fuel is considered a  high priority and that the current goal is to qualify  the fuel in a similar manner to already-qualified  fuels. H e  further stated that the intention was to  convert all US research reactors to monolithic fuel  by 2012.

6. REPLACEMENT RESEARCH REACTOR

The Preliminary Safety Analysis Report (PSAR) for  the Replacement Research Reactor (RRR) was  written and accepted by the regulator on the basis of  design and operation using LEU 4.8 gU/cm3 U$i2  fuel. INVAP, the suppliers of RRR, are contracted  to provide the first 2 cores (32 fuel assemblies) of  silicide fuel. These will be used for commissioning  and initial operation. A further two cores of silicide  fuel with be purchased by ANSTO to cover reactor

operation up to the availability of UMo fuel.  ANSTO has in place arrangements for the

reprocessing and disposal of this RRR spent silicide  fuel. It is therefore not restricted by the May 2006  deadline that is critical to the operators of the  Swedish R2 reactor.

To date,

UMo fuel development has mainly  concentrated on U7Mo with a density of 6 gU/cm3.  This fuel is planned to have virtually identical  behaviour to 4.8 gm/cm3 silicide fuel, in order to  minimise licensing problems with the changeover.  In view of the expccted acceptability of UMo fuels  for reprocessing and the possibility of a future

development path for increased fuel density and  hence neutron performance, it is still planned that the  RRR will change

to using UMo fuel shortly after the  fuel is qualified.

7.

4

b)

c>

d)

8.

1.

2.

3.

4.

5 .

6.

CONCLUSIONS

The US RERTR program has continued its push  to minimise worldwide use of LEU and this is  now planned to extend to the high powered US  and Russian research reactors that were

previously considered unsuitable for LEU fuels.

There have been delays in the qualification of  dispersion UMo fuel, with qualification by the  US NRC now not expected before 2006.

”Monolithic” UMo with a density of up to  16.7 gU/cm’ has shown potential as a fuel and  research programs are in place to study it

further.

ANSTO’s RRR will start up with conventional  4.8 gU/cm3 silicide LEU fuel and will operate  with silicide fuel beyond the initial two cores  due to the delay in UMo qualification.

REFERENCES

Finlay, M. R. and Ripley, M. I. (2001) A  new fuel for research reactors. ANA 2001:  Fourth Conference on Nuclear Science and  Engineering in Australia, 24-25 Oct 2001  Sydney, Australian Nuclear Association.

Gay, A. and Beliers, M. (1997) Reprocessing  Ability of High Density Fuels for Research  and Test Reactors. Proc. ”20th Int. Meeting  on Reduced Enrichment for Research and  Test Reactors (RERTR)”, 5-10 Oct 1997,  Jackson Hole WY

Hofman, G. L. and Meyer, M. K. (2002)  Progress in irradiation performance of  experimental uranium-molybdenum  dispersion fuel. Proc. “24th Int. Meeting on

Reduced Enrichment for Research and Test  Reactors (RERTR)”, 3-8 Nov 2002, Bariloche

Kim, K. H., Lee,  D. B., Kim, C. K., Kuk, I.

H., and Hofman, G. L. (1996) Development  of high-loading U-Mo alloy fuel by

centrifugal atomization. Proc. 1996 Int.  Meeting on Reduced Enrichment for Research  and Test Reactors (RERTR)”, 6-11 Oct 1996,  Seoul.

Nuclear Regulatory Commission, Office of  Nuclear Reactor Regulation, Washington  DC.  (1988) Safety evaluation report related to the

evaluation of low-enriched uranium silicide-  aluminum dispersion fuel for use i n  non-  power reactors. NUREG- 13 13, 1 19p.  Pasqualini, E. E., Helzel Garcia, J., Lopez,

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M., Cabanillas, E., and Adelfang, P. (2002)  Powder production of U-Mo alloy; HMD  process. Trans. 6th Int. Topical Meeting on  Research Reactor Fuel Management ( R M ) .

17-20 Mar 2002, Ghent. European Nucl. SOC.

7. Snelgrove, J., Lemoine, P., Adelfang, P., and  Arkhangelsky, N. (2003) Qualification and  licensing of U-Mo fuel. Trans. 7th Int.

ALLOY ALLOY TYPE

U metal

UA1,-AI Cast

UA1,-A1 Dispersion

U3Si2-AI Dispersion

UMo-AI Dispersion

Topical Meeting on Research Reactor Fuel  Management (RRF’M) 9-12 Mar 2003,

Aix-

en-Provence European Nucl. SOC.

8. Travelli, A. (2002) Status and Progress of the  RERTR Program in the Year 2002. Proc.  “24th International Meeting on Reduced  Enrichment for Research and Test Reactors

(RERTR)” 3-8 Nov 2002, Bariloche

PHASE DENSITY OF URANIUM (g/cm’)

IN PHASE IN FUEL MEAT

a 19.0 19.0

UA14 6.1 I .7

UA14 6.1 2.4

UAli 6.8

UA12 8.1

UiSi2 12.2 4.8

UISi 15.2

U-6Mo 16.7 6-8

U-8Mo 16.0

U- 1 OM0 15.3

Program French

Fabricator CERCA

Reactor OSIRIS (Fr)

Number  of elements 2

RERTR RERTR CNEA

RWXT CNEA CNEA

HFR (Neth) HFR (Neth) RA-3 (Arg)  2 2 2

121

Figure 1 .  Proposed schedule for qualification and generic approval of plate-type fuels as of March 2003. 171.

r

Experiment RERTR- 1 RERTR-2 RERTR-3 RERTR-4 RERTR-5

Reactor ATR (US) ATR (US) ATR (US) ATR (US) ATR (US)

Number of small plates(") 20 I 32 22 I32 43 I47 25 I 32 27 I32

Gr, At Gr, At At Mach, At Mach, At

Enrichment (% %) 19.5 19.5 19.5 19.5 19.5

6.0, 8.0 Fuel loading (g Ulcm') 4.0 4.0 8.5 6.0, 8.0

Heat flux at BOL (Wlcm') 55 55 390 210 3 20

Cladding temperature ("C) 65 ~ _ _  65 150 ~ 130 175

Coolant velocity (m/s) 7.6 7.G 6.0 4.0 4.0

Status of experiment Complete Complete Complete in PIE i n  PIE

Number of irradiation cycles 2 8 2 7 3

Total duration (days) 94 232 48 230 116

Average burnup at EOL (% 40 70 40 80 50

FUTURE

IRIS2

French Elements  - RERTR Elements CNEA Element

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Experiment IRIS1 UMUS

Reactor OSIRIS (Fr) HFR (Neth)

Number of full-sized plates 3 2 1 2

U-Mo powder Ground Ground

Enrichment (% 235U) 19.7.5 19.7.5 t 35

Fuel loading (g Ulcm’) 8.0 8.0

Max. heat flux at BOL (Wlcm’) 120 170

Figure 2. Optical micrograph of atomised U-1OMo fuel irradiated at -120Â°C (beginning of life temperature).  Note the extensive interaction between the fuel particles (<12Oym diameter) and the surrounding  aluminium matrix [3].

IRIS2 FUTURE

OSIRIS (Fr) BR2 (B)

4 2

Atomized Atomized  19.75 19.7.5

8.0 8.0

230 340

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