Fuel bundle irradiation to high burnup

To investigate fuel behavior at high burnup with regard to fission gas release, fuel swelling and sheath material behavior, fuel bundle integrity and power ramp performance; it was decided to irradiate few high burnup fuel bundles of different types including natural uranium bundles in the operating PHWRs. Short length fuel bundles and on-power refueling provision in PHWRs provides flexibility to use variety of fuel types and consequently permits optimum use of fuel in the reactor. Following paragraphs cover the alternative fuel design concepts in use or under consideration for use in Indian PHWRs.

3.1 Natural uranium fuel bundles

It was planned to prolong irradiation of few natural uranium bundles present in core, to higher burnups. For this purpose natural uranium bundles in two channels (H-13 and O-08 of KAPS-2^a core) were selected and fuel irradiated upto a discharge burnup of 20 000 MW•d/Te U. Bundles in these channels were irradiated to about 1100 FPD which is about 35% more compared to normal irradiation time[2]. Performance of these channels was assessed by Delay Neutron Monitoring (DNM) system and by comparing the measured Channel Outlet Temperature (COT) with the estimated values by the fuel management code TRIVENI. The variation of DN counts of these channels during this period and also during refueling of these channels indicated successful performance of the fuel bundles including the power ramps seen during operation at high burnup. Fourteen fuel bundles of these two channels have seen burnups more than 15 000 MW•d/Te U, the maximum being 22300 MW•d/Te U.

a Kakrapar Atomic Power Station Units 1 and 2.

3.2 Thorium bundles

It was planned to use Thorium bundles for flux flattening in the initial core such that the reactor can be operated at rated full power in the initial phase. The Thorium bundle is a 19–element fuel bundle with Thorium dioxide as fuel in pellet form [6]. The pellet shapes used are both flat and single dish type. As shown in Figure 3 the bundle power of these bundles gradually increase with irradiation exposure time due to production of fissile isotope U²³³ [5]. The fuel element Thermo-mechanical analysis was carried out for elements operating on such an envelope. The elements are designed for a peak Linear Heat Rating of 57.5 KW/M and burnup of 15 000 MW•d/Te Th. The Thorium dioxide pellet specification was evolved which consists of chemical content, density, shape specifications. High density ThO2

pellets suitable for PHWR were developed at BARC^a [7]).The fuel element axial and radial gaps have been suitably specified. By carrying out minor modification in their bearing pad positions, proper identification of these bundles was provided. The fuel bundles were fabricated by NFC^b.

Initially four lead thorium bundles were irradiated in MAPS -1^c reactor during the eighties.

Subsequently, 35 Thorium bundles have been used as a part of initial charge fuel in the 220 MW(e) PHWRs for flux flattening in the initial core such that the reactor can be operated at rated full power in the initial phase. These bundles are distributed throughout the core in different bundle locations, both in the high power and low power channels. This loading was successfully demonstrated in KAPS–1 and subsequently adopted in the initial reactor loading of KAPS–2, KGS–1 and 2^d and RAPS 3 and 4^e. So far, 232 thorium dioxide bundles have been successfully irradiated in different reactors [8]. The maximum fuel bundle power and burnups seen are 408 kW and 13 000 MW•d/TeTh respectively.

These bundles withstood the power ramps normally experienced in reactor while the typical power envelope of thorium fuel is such that power increases with irradiation.

FIG. 3. Natural uranium and MOX–7 fuel bundles.

3.3 MOX–7 bundles

MOX–7 bundle design has been evolved, which is a 19–element cluster, with inner seven elements having MOX pellets consisting of Plutonium dioxide mixed in natural uranium dioxide and outer 12 elements having only natural uranium dioxide pellets[9]. Figure 4 shows typical MOX fuel bundle.

Based on detailed studies, an optimised loading pattern and refuelling scheme has been evolved for loading initially 50 lead MOX bundles in an existing operating reactor. The LHR for these elements are maintained similar to the 19–element natural uranium bundle outer elements. Subsequently For initial trial irradiation 50 number of MOX–7 bundles have been fabricated by BARC and NFC. Unlike natural uranium bundles, elements of these bundles are seal welded by TIG welding. The pellets are of

a Bhabha Atomic Research Centre, Mumbai, India.

b Nuclear Fuel Complex , Hyderabad, India.

c Madras Atomic Power Station Units 1 and 2.

d Kaiga Generating Station Units 1 and 2.

e Rajasthan Atomic Power Station Units 3 and 4.

single dish pellets. Fifty number of MOX bundles were loaded in the KAPS–1 reactor in different locations in the year 2004 [9]. In each refueling four MOX, bundles are loaded in the bundle locations 5 to 8 of the channel. In few channels MOX bundles were loaded in 4^th location and subsequently shuffled to 8^th location in the same channel. In order to obtain higher bundle power production from MOX bundles and achieve desired burn up at the earliest, bundles producing about 300 KW in low power channels like K–03 were recycled to central channels at a burnup of about 2000 MW•d/Te HE.

These bundles successfully withstood the power ramps. Forty six bundles, after 16 months of irradiation and accumulating a burnup of about 11 000 MW•d/Te HE, have been discharged as a part of normal refuelling. Four bundles were irradiated to a burnup of 20 000 MW•d/Te HE in one chanel.

The performance is good. The DN counts of these channels were steady, indicating good fuel performance of these bundles. All the discharged bundles were sniffed in spent fuel bay and found nondefective.

1 00

0 5 000 1 0000 1 5 000

Bundle Power ^UO2

MOX

Thoriu m

FIG. 4. Bundle power envelopes for different fuel types.

3.4 SEU fuel bundles

India has also started analysis and design works for PHWRs using slightly enriched uranium. This offers higher burnup and consequently less annual fuel requirement and spent fuel inventory. R&D works for design of full core of PHWRs with SEU are completed. Fuel design issues in respect of 0.9% to 1.1% U–235 isotopic content have been reviewed. The core average discharge burnup increases to 14 000 MW•d/Te U with 1.1% enrichment, thermo mechanical analysis of 19–element fuel bundle up to 25 000 MW•d/Te U burnup was carried out for 0.9% EU envelope. The void volume inside the fuel element is increased by increasing the dish depth to maintain low internal fission gas pressure. To accommodate Fuel swelling at higher burnups upto 25 000 MW•d/Te U, fuel (UO2) density is maintained at about 96 to 97% theoretical density. Few SEU bundles were fabricated by NFC and are loaded in MAPS–2 unit recently for trial irradiation. The channels in which EU bundles are loaded are kept under watch and the DN Counts of these channels are closely observed. The performance of these bundles in core is satisfactory so far.

4. Conclusions

Indian nuclear power programme is based on optimum utilisation of available uranium and thorium resources in the country. The fuel designs and fuel usage strategies are evolved based on this objective. As part of this, high burnup fuel design and irradiation studies have been takenup. Few natural uranium bundles in KAPS–2 were irradiated upto 3.5 years. Also the MOX bundles in KAPS–

1 were irradiated upto 2.5 years. This experience gives confidence that the high burnup fuel behaviour in core is satisfactory without any deterioration due to corrosion and irradiation embrittlement. The irradiation performance of both the graphite coated natural U and MOX bundles in the reactors gives the confidence that the graphite coating works at the burnups experienced by them and the bundles can withstand the power ramps due to neighboring channel fuelling.

REFERENCES

[1] BHARDWAJ, S.A., et al, Fuel design evolution in Indian PHWRs, International Symposium on improvements in water reactor fuel technology and utilization, IAEA, Stockholm, (1986).

[2] BHARDWAJ, S. A., Kumar, A.N., Prasad, P.N., Ravi, M., Fuel performance, design and development in Indian reactors, International Conference on CANDU fuel, (2003).

[3] TRIPATHI, R.M., PRASAD, P. N., DWIVEDI, K.P., Thermo-mechanical parametric studies on high burnup bundles for 220 MW(e) PHWRs , CQCNF–2009, (2009).

[4] FUDA MOD–2 manual,”NPC–500/DC/37000/08–Rev–0”, (1996)

[5] BHARDWAJ, S.A., SIVASUBRAMANIAN, S., LEE, S.M., Current status and future possibilities of thorium utilisation in PHWRs and FBRs, Annual Conference of Indian Nuclear Society–2000, Mumbai, (2000).

[6] DAS, M., et al, Design and irradiation experience with torium bsed fuels, Thorium Cycle Activities in India, Bombay (1990).

[7] VIJAYAEAGHAVAN, R., et al., Development and fabrication of high density Thoria pelltets, Presented at the Thorium cycle activities in India, (1990).

[8] KUMAR, A.N., et al, Experience with Kakrapar atomic power plants with Thorium loading”, Annual Conference of Indian Nuclear Society–2000, Mumbai, (2000).

[9] BAJAJ, S.S., PRASAD, P.N., RAVI, M., KUMAR, A.N., Fuel design and performance in Indian reactors, The International Conference on Characteristics of Quality Control of Nuclear Fuels, Hyderabad, (2003).

[10] PARIKH, M.V., et al., Status report on MOX–7 bundles loaded in KAPS–1 core, (2008), NPCIL Internal Report.

Research on Defects of Pressure Resistance Welding for HWR Fuel