XA9744131 AECL's EXPERIENCE IN MOX FUEL FABRICATION AND IRRADIATION

F.C. DIMAYUGA Laboratories, Atomic Energy of Canada Limited, Chalk River, , Canada

Abstract

Atomic Energy of Canada Limited's mixed-oxide (MOX) fuel fabrication activities are conducted in the Recycle Fuel Fabrication Laboratories (RFFL) at the Chalk River Laboratories. The RFFL facility is designed to produce experimental quantities of CANDU* MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions. About 150 bundles, containing over three tonnes of MOX, were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is underway. The initial campaign will consist of the production of thirty-eight 37-element (U, Pu)O2 bundles containing 0.3 wt. % Pu in Heavy Element (H.E.) destined for physics tests in the zero-power ZED-2 reactor. An overview of AECL's MOX fuel irradiation program will be given. Post-irradiation examination results of (U,Pu)O2 bundles irradiated to burnups ranging from 18 to 49 GWd/te H.E. in the Nuclear Power Demonstration reactor are highlighted. The results demonstrate the excellent performance of CANDU MOX fuel to high burnup, at power ratings up to 45 kW/m. The paper also outlines the status of current MOX fuel irradiation tests, including the irradiation of various (U, Pu)O2 and (Th, Pu)O2 bundles.

1. INTRODUCTION

AECL's MOX fuel program was started more than thirty years ago. The program consisted of two components:

• irradiation testing and post-irradiation examination (PIE) of various types of MOX fuel, with the

objective of studying the general behaviour of these types of fuel in comparison with natural UO2 fuel; and, • development of MOX fuel fabrication technology.

Irradiation testing and PIE of MOX fuel is still continuing at AECL; it has gone from multi- element to multi-bundle demonstration testing. This paper discusses recent PIE results of (U,Pu)O2 bundles irradiated in the Nuclear Power Demonstration (NPD) reactor, and outlines the status of current

MOX fuel irradiation tests, including the irradiation of various (U,Pu)O2 and (Th,Pu)O2 bundles in the National Research Universal (NRU) reactor.

AECL's MOX fuel fabrication experience is summarized, including an update on the current fabrication campaign.

*CANDU: CANada Deuterium ; registered trademark.

373 2. MOX FUEL FABRICATION AT AECL

2.1 Recycle Fuel Fabrication Laboratories

Various forms of fuel containing have been handled by AECL at its Chalk River Laboratories since 1960. Research activities were carried out in glove boxes between 1960 and 1970, including the development of MOX fuel fabrication technology, measurement of physical properties, production of fuel samples for experimental irradiation, etc. In 1970, a decision was made to re-model the plutonium laboratory and install new facilities, to focus on MOX fuel fabrication technology. Installation of the new facilities was complete by 1975 [1], The facility, collectively referred to as the Recycle Fuel Fabrication Laboratories (RFFL), is designed to produce experimental quantities of MOX fuel for reactor physics tests or demonstration irradiations.

Subject to special precautions due to the presence of Pu (e.g., essentially all operations are done inside glove boxes), the processes employed in the RFFL follow conventional natural UO2 practice. The fabrication line was designed for the production of sealed individual fuel elements, starting from UO2 or ThO2 powders as the major component and PuO: or ^UO; as the minor component.

The fabrication process adopted in the RFFL is outlined in Fig. 1. Weighed amounts of the starting oxide powders are mixed either in single-stage or double-stage blending, the latter being used for more dilute mixtures to achieve better homogeneity. After blending, the MOX powder is pre-pressed using an isostatic press, to convert the mixed powder into compacts, which are, in turn, fed into a granulator. The resulting free-flowing granules are then suitable for final pressing into green pellets using a single-cavity hydraulic press.

The green pellets are then loaded into a batch furnace, where sintering is done in a dilute hydrogen cover gas. Sintered pellets are then centreless ground to a specified diameter and surface finish. The pellets are washed and then dried in warm air. Acceptable pellets are loaded into empty sheaths that already have one endcap welded and all appendages brazed in place (these sub-assemblies are supplied by commercial fabricators). The second endcap is welded to the loaded sheath using a tungsten inert gas (TIG) welding system. The sealed elements are then helium leak-tested, scanned for surface alpha contamination, and assayed by neutron interrogation. Following assay, the elements are ready for bundle assembly.

2.2 Fabrication Campaigns in the RFFL

Extensive commissioning and test production programs using natural UO2 were implemented in the RFFL. This enabled several modifications to, and optimization of, various process and safety-related equipment prior to the introduction of Pu in the fabrication line in 1979.

From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions [1]. As listed in Table I, the first campaign consisted of (U, Pu)O2 fuel with 0.5 wt.% Pu in H.E. Later, about 1.3 Mg of (Th,Pu)O fuel elements were produced, containing up to 2.3 wt.% Pu in H.E. The last campaign was particularly challenging, since it involved the fabrication of 1350 elements containing 1.4 wt.% ^U in H.E. (due to the presence of232!!, which has a gamma-active daughter) [2].

The fuel elements and bundles produced in the RFFL were used for test irradiations in NRU, and for physics tests in the zero power ZED-2 reactor. About 150 bundles, containing over three tonnes of MOX were fabricated in the RFFL before operations in the facility were suspended. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained.

Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is underway. The pre-operational testing (using inactive material) is scheduled for mid-1995, with regulatory approval to

374 resume MOX operations anticipated by late 1995. The initial campaign will consist of the production of thirty-eight 37-element (U,Pu)O: bundles containing 0.3 wt.% Pu in H.E., destined for physics tests in ZED-2.

Product Analysis MOX Fuel Production and QA/QC

Sample to Analysis

Return Repackaged PuO2 to Storage

Sample to Analysis Granulate & Sieve Optional Operations

Master-Mix Pre-weigh

Sample to Analysis

Waste Management & Recycle

Sample for Recover Metallography and Grinding Fines Auto radiography

Sample for Density Measurement

Ultrasonic Wash & Dry

100% visual H Vacuum Degas Inspection

100% Diameter ft Sample to visual Inspection Analysis

Weld Second End-Cap Immobilize >,,. Liquids 100% He Leak Test, -4 a Contamination Store Finished "\ Transfer to Waste Check, Dimensions, Elements Treatment Centre Assay

FIGURE 1. RFFL Fabrication Process Flowsheet.

375 TABLE I Fuel Produced in the RFFL.

Experiment DATE FUEL TYPE QUANTITY DP-12 1977-78 Natural UO2 50, 19-element bundles BDL-419 1979-80 (U, Pu)[email protected]%Pu 15, 36-element bundles BDL-422 1981-83 (Th, Pu)O,@ 1.75% Pu 6, 36-element bundles BDL-430 1982 Natural ThO2 1, 36-element bundle WR1-1012 1982 (Th, U)[email protected]%Pu 2, 21 -element bundles 1982 (Th, Pu)O2 @ 2.3% Pu 2, 21-element bundles WRl-1010 1982-85 (Th, Pu)O2 @ 2.3% Pu 1332 elements BDL-432 1986-87 (Th, U)O2@ 1.4%U-233 1350 elements

3. AECL's RECENT IRRADIATION EXPERIENCE (NPD-40)

3.1 The NPD-40 Experiment

The NPD-40 experiment was a demonstration-scale irradiation test involving six bundles containing (U,Pu)O2 fuel in NPD. The objective of the experiment was to show that MOX fuel could operate successfully to extended burnup, under typical CANDU operating conditions. Subsequent to the NPD irradiation, several bundles were power-ramped in NRU.

3.2 Fuel Design and Fabrication

The six bundles were of the standard 19-element design, but contained two different types of non- standard fuel pellets from two different suppliers, and two sheath wall thicknesses. As listed in Tables II and HI, three bundles had normal-(CANDU)-density hollow pellets (supplied by Aktiebolaget Atomenergi, Sweden) contained in thin-wall, collapsible Zircaloy-4 sheaths, while the other three had low-density solid pellets (supplied by BelgoNucleaire, Belgium) contained in thick-wall, free-standing Zircaloy-4 sheaths. Both pellet designs considered the effects of in-reactor operation:

• the low-density pellets had more internal voidage to accommodate pellet swelling during operation. In-service densification was accounted for by using free-standing sheaths; and

• the hollow design of the normal-density pellets served to reduce pellet thermal swelling, due to its lower average operating temperature.

The pellets were produced by both suppliers using normal fabrication methods; i.e., mixing and blending, pre-pressing, granulation, final pressing, sintering and grinding. In addition, the low-density pellets were degassed immediately prior to loading into the sheaths.

Zircaloy-4 sub-assemblies (i.e., appendaged sheaths with one endcap welded in place, and with graphite CANLUB coating on the sheath inside surface) were obtained from General Electric (GE) Canada, a commercial fuel supplier. Suitable endcaps for final-closure TIG welding were also supplied with the sub-assemblies. Glove boxes were used to load measured stacks of pellets and plenum inserts into die element sub-assemblies, and to fit the TIG endcaps. Final closure welds were then performed using a TIG welder.

Kits of 19 finished elements were sent to GE Canada for bundle assembly welding. The bundles were then returned to Chalk River Laboratories for inspection prior to irradiation in NPD.

376 TABLE II. Fuel Design Data for Bundles KA, KB and KC of Experiment NPD-40 (Dimensions in mm unless otherwise stated)

Bundle Design 19-Element Geometry

Overall Length 495.0 Overall Diameter 81.6

Element Design

Overall Length 492.6 Internal Diameter 14.44 Wall Thickness 0.38 Sheathing Material Zircaloy-4 (U,Pu)O2 Weight/Element (g) 740.0

(U.Pu) Oxide Pellets

Manufacturer Aktiebolaget Atomenergi Density (Mg/m3) 10.5 Outside Diameter 14.2 Central Hole Diameter 2.54 Length 14.6 Enrichment 3.33 wt% (Pu + U) Fissile/HE Weight (U,Pu)O2/Bundle (kg) 14.06

TABLE III. Fuel Design Data for Bundles KD, KE and KF of Experiment NPD-40 (Dimensions in mm unless otherwise stated) Bundle Design 19-Element Geometry

Overall Length 495.0 Overall Diameter 81.6

Element Design

Overall Length 492.6 Internal Diameter 13.77 Wall Thickness 0.65 Sheathing Material Zircaloy-4 (U,Pu)O2 Weight/Element (g) -685

(U.Pu) Oxide Pellets

Manufacturer Belgonucleaire Density (Mg/m3) 10.2 Diameter 13.67 Length 14.68 Enrichment 3.0wt%Puin(U + Pu) Weight (U,Pu)O2/Bundle (kg) 13.0

377 3.3 NPD Irradiation

Three bundles were loaded into NPD in 1973 January, and the other three were loaded in 1973 December. For most of the NPD irradiation, the bundles operated in pressurized heavy-water coolant at about 260°C and about 7 MPa pressure. Coolant was maintained at pH 10 with LiOH, and conductivity controlled to less than 1.5 mS/m.

As shown in Table IV, most bundles had an average element burnup of 14 GWd/te with the outer elements having a maximum burnup of 18 Gwd/te. (1 Gwd/te = 24 MWh/kg). The exception is bundle KB, which was operated to an average element burnup of 45 GWd/te, or an outer-element burnup of 49 Gwd/te. Figures 2-7 show the power history of each of the bundles; peak linear ratings ranged from 40 to 50 kW/m at beginning-of-life, decreasing to about 15-25 kW/m at end-of-life.

3.4 NRU Power Ramp Tests

Following the NPD irradiation, three bundles (one with normal density pellets, bundle KA, and two with low-density pellets, bundles KE and KF) were tested in NRU to observe MOX fuel performance during power ramps. After a few hours of low-power soak (about 20 kW/m linear rating), the bundles were ramped to 50 kW/m. This power ramp produced some defects in bundle KA. Bundles KE and KF survived and operated at 50 kW/m for 10 days. A further increase in power to about 70 kW/m resulted in a defect in one of the bundles, bundle KE.

3.5 Post-Irradiation Examination

All bundles were visually examined after the NPD irradiation. No peculiar findings were observed. Several bundles were measured for their element dimensional changes. In general, the elements showed slight diametral contraction from the unirradiated dimension. For the thin-wall, collapsible sheaths, measurements indicate a decrease in diameter up to a maximum of about 0.7%. For the thick-wall, free-standing sheaths, measurements also show a decrease in element diameters up to a maximum of about 0.3%.

After the power ramp in NRU, elements from bundle KA still indicated slight diametral contraction. After power ramping to 70 kW/m, outer elements of bundles KE and KF showed diametral expansions of about 1 %. Inner elements of both bundles continued to show a diametral contraction.

Internal gases were extracted, measured and analyzed by mass spectrometry for all bundles except KD. As listed in Table V, the non-ramped bundles (KB and KC) had relatively low fission-gas release in the range from 2 to 4% of that produced for the outer elements. However, those elements that were ramped to powers above 60 kW/m (KE and KF) experienced high gas release in the order of 40 to 50% of that produced. Bundle KA, which was ramped to about 50 kW/m, had about 11 % fission-gas release from the outer elements.

TABLE IV. NPD-40 Element Burnup (GWd/te HE.)

BUNDLE Centre Element Inner Element Outer Element Bundle Average KA 7.1 9.2 17.5 14.2 KB 35.4 38.8 49.2 45.0 KC 7.1 9.6 17.5 14.6 KD 7.1 9.2 16.2 13.3 KE 7.1 8.8 15.4 12.9 KF 7.9 10.0 16.7 14.2

378 NPD-40 BUNDLE KA POWER HISTORY

SO 100 150 200 250 300 3S0 400 450 500 ELEMENT BURNUP (MW.h/kgH.E)

• Outer element - Inner element - Centre element

FIGURE 2. NPD-40 Bundle KA Power History.

NPD-40 BUNDLE KB POWER HISTORY

100 200 300 400 500 600 700 800 900 1000 1100 1200 ELEMENT BURNUP (MW.h/kgH.E)

Outer element Inner element - Centre element

FIGURE 3. NPD-40 Bundle KB Power History.

379 NPD-40 BUNDLE KC POWER HISTORY

50 100 150 200 250 300 350 400 450 500 ELEMENT BURNUP (MW.h/kg H.E)

CENTER INNER OUTER

FIGURE 4. NPD-40 Bundle KC Power History.

NPD-40 BUNDLE KD POWER HISTORY

100 150 200 250 300 350 400 ELEMENT BURNUP (MW.h/kg H.E)

Centre element Inner element Outer element

FIGURE 5. NPD-40 Bundle KD Power History.

380 NPIMO BUNDLE KE POWER HISTORY

70 T

50 100 150 200 250 300 350 400 ELEMENT BURNUP (MW.h/kgH.E)

Outer element Centre element Inner element

FIGURE 6. NPD-40 Bundle KE Power History.

NPD-40 BUNDLE KF POWER HISTORY

70 T

50 100 150 200 250 350 400 ELEMENT BURNUP (MW.h/kgH.E)

' Outer element - Centre element - Inner element

FIGURE 7. NPD-40 Bundle KF Power History.

381 TABLE V. Xenon Release for NPD-40 Bundles

Bundle-Element Calculated Volume of Measured Volume of Percentage of Xenon Xenon Formed (mL) Xenon (mL) Released (%) KA-01 780 86.9 11.1 KA-02 780 86.3 11.1 KA-13 401 13.3 3.3 KB-02 955 18.8 2.0 KB-04 956 38.9 4.1 KB-07 956 31.6 3.3 KB-10 954 40.6 4.3 KB-13 748 12.0 1.6 KB-16 749 9.9 1.3 KC-02 331 7.8 2.4 KC-13 178 0.3 0.2 KE-07 270 125.8 46.6 KE-08 271 129.3 47.7 KE-09 271 122.6 45.2 KE-10 271 127.6 47.0 KE-11 271 127.9 47.2 KF-02 293 142.2 48.5 KF-10 294 138.4 47.1 KF-17 173 43.4 25.1

NOTE: In the 19-element bundle geometry, elements 1-12 are located in the outer ring, elements 13-18 are in the intermediate ring, and element 19 is the centre element.

Metallographic examination showed that the microstructures had isolated regions of porosity and metallic fission-product precipitates (Fig. 8). These were attributed to the presence of high-fissile-content particles in the as-sintered pellet (Fig. 9), which resulted in zones of increased fission density and localized increase in temperature during operation. It was postulated that most of the fission-gas release was from these zones of initial high fissile Pu content.

PIE measurements indicated that elements containing low-density pellets had thicker CANLUB coatings than those containing normal density pellets (maximum thicknesses observed were 40 and 8 pm, respectively). There is no clear reason for this, although one possibility is the difference in as-fabricated coating thickness (the bundle kits were fabricated at different periods during 1972). Another possibility is the experimental error involved in determining the amount of coating retained. The thicker CANLUB coatings may have added to the benefits of low-density pellets resulting in the observed superior performance of these bundles during power-ramp tests in NRU.

3.6 Conclusions of NPD-40

This test demonstrated the excellent performance of dry-blended MOX fuel to high burnups (maximum of 49 GWd/te) at power ratings up to 45 kW/m. Dimensional changes and fission-gas release were minimal, and there were no indications of defects during the irradiation. There were no significant differences in the behaviour of the two fuel types (i.e., low-density pellets in thick-wall, free-standing sheaths and normal CANDU density annular pellets in thin-wall, collapsible sheaths).

382 HGURE 8. Microstructure of element KB-01 showing porous zone with metallic precipitates.

FIGURE 9. Microstructure of an archive pellet showing second-phase particles after etching.

383 4. CURRENT MOX FUEL IRRADIATIONS AT AECL

There are currently two MOX fuel irradiation tests in NRU: experiment numbers BDL-419 and -422. Both experiments involve demonstration of the performance of fuel bundles fabricated in the RFFL in comparison with UO; fuel. The bundles are of the standard 37-element Bruce design. BDL-419 contains (U,Pu)O2 fuel with 0.5 wt% Pu in H.E., and BDL-422 contains (Th,Pu)O: fuel with 1.4 wt% Pu inH.E.

An initial set of six BDL-419 bundles was successfully irradiated in NRU up to burnups of 11-22 GWd/te at peak linear ratings of 35-65 kW/m. PIE of two bundles from this batch was recently completed, and documentation of the results is underway. Preliminary results indicate that the fuel performed very well, with fission-gas releases of less than 13%. A further set of six bundles from the same fabrication campaign are currently under irradiation in NRU. Outer element burnups are in the range of 7-12 GWd/te. In light of the excellent performance of the first set of bundles, the target burnup for the second set is under review, and will likely be increased to the range of 30-40 GWd/te.

BDL-422 consists of irradiation testing of six (Th,Pu)O: fuel bundles in NRU. Outer-element burnups are presently at 17 GWd/te with a target level of 42 Gwd/te.

5. SUMMARY

AECL's MOX fuel fabrication activities are conducted in the RFFL at Chalk River Laboratories. The RFFL facility is designed to produce experimental quantities of CANDU MOX fuel for reactor physics tests or demonstration irradiations. From 1979 to 1987, several MOX fuel fabrication campaigns were run in the RFFL, producing various quantities of fuel with different compositions. In late 1987, the RFFL was placed in a state of active standby, a condition where no fuel fabrication activities are conducted, but the monitoring and ventilation systems in the facility are maintained. Currently, a project to rehabilitate the RFFL and resume MOX fuel fabrication is underway. The initial campaign will consist of the production of thirty-eight 37-element (U, Pu)O; bundles containing 0.3 wt. % Pu in H.E., destined for physics tests in the zero-power ZED-2 reactor.

This paper has provided an overview of AECL's MOX fuel irradiation program. Irradiation testing and PIE of MOX fuel is continuing at AECL; it has gone from multi-element to multi-bundle demonstration testing. PIE of (U,Pu)O2 bundles irradiated to burnups ranging from 18 to 49 GWd/Mg H.E. in NPD has demonstrated the excellent performance of MOX fuel to high burnup, at power ratings up to 45 kW/m, including power ramp testing. Dimensional changes and fission-gas release were minimal, and there were no indications of defects during the irradiation.

The status of current MOX fuel irradiation tests in NRU has been outlined, including the irradiation of various (U, Pu)O2 and (Th, Pu)O2 bundles to extended burnup.

ACKNOWLEDGEMENTS

The author acknowledges several people whose work has formed the basis for this paper, particularly those who were involved in the various fabrication campaigns in the RFFL and the NPD-40 experiment. Among them are T.J. Carter, R.J. Chenier, R.R. Elder, M.F.M. Primeau, D.H. Rose, R. Behnke, D.L. Bruneau, I.A. Lusk, K.D. Vaillancourt, L. Norlock and R. Stack. Thanks are also given to M.R. Floyd and D.S. Cox for their thorough review of this paper and the helpful suggestions they have provided.

384 REFERENCES

[1] CARTER, T.J., "The Recycle Fuel Fabrication Laboratory at Chalk River", Proc. Int. Conf. CANDU Fuel, 1986, Chalk River, Ontario, Canada, (HASTINGS, I.J., Ed.), Canadian Nuclear Society, Toronto (1986), pp. 168-176. [2] CARTER, TJ. and JONES, R.T., "The Large Scale U-233, Thoria Reactor Physics Experiment at Chalk River Nuclear Laboratories", Proc. Second Int. Conf. CANDU Fuel, 1989, Pembroke, Ontario, Canada, (HASTINGS, I.J., Ed.), Canadian Nuclear Society, Toronto (1989), pp. 389- 397.

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