Post Irradiation Examination of Thoria-Plutonia Mixed Oxide Fuel in Indian Hot Cells

Transcription

1 Post Irradiation Examination of Thoria-Plutonia Mixed Oxide Fuel in Indian Hot Cells J. L. Singh, Prerna Mishra, J. S. Dubey, H. N. Singh, V. D. Alur, P. K. Shah, V. P. Jathar, A. Bhandekar, K. M. Pandit, R. S. Shriwastaw, P. M. Ouseph, Nitin Kumawat, P. M. Satheesh, G. K. Mallik and Arun Kumar Post Irradiation Examination Division Nuclear Fuels Group Bhabha Atomic Research Centre Trombay, Mumbai Abstract Mixed oxide (MOX) fuel clusters containing ThO 2 +4%PuO 2, and ThO %PuO 2 fuel pins were irradiated in the pressurized water loop of the Indian research reactor CIRUS, to burn up in the range of 20 GWd/T(HM). The ThO 2 +4%PuO 2 fuel elements had free standing cladding made of Zircaloy-2 and the ThO %PuO 2 had collapsible Zircaloy-2 cladding. The fuel clusters had performed well during irradiation with no apparent indications of failure. The techniques used for the post irradiation examination (PIE) of these fuels in the hot cells included visual examination, fuel pin diameter measurements, leak testing, gamma scanning, gamma spectrometry, ultrasonic testing, eddy current testing, ceramography, metallography, beta gamma autoradiography and measurement of released fission gases. Micro hardness measurement of cladding and evaluation of mechanical properties using ring tension test were also carried out. This paper elaborates on the techniques and the results of the PIE carried out on ThO 2 +4%PuO 2 fuel. 1. Introduction India has about four times more thorium resources than uranium. Utilization of thorium for large scale energy production is a major goal in the three stage nuclear power programme [1]. Hence, research and development in fabrication, characterization and irradiation testing of thoria based fuels is necessary. A six pin cluster, consisting of ThO 2-4% PuO 2 fuel pellets, had undergone irradiation testing in the pressurized water loop (PWL) of CIRUS thermal reactor up to a burn-up of 18.5 GWd/t. Post irradiation examination (PIE) of the fuel pins from the cluster was carried out at BARC hot cells facility. Non-Destructive Examination of the fuel pins from the cluster has been carried out by visual examination, fuel pin diameter measurements, leak testing, gamma scanning, gamma spectrometry, ultrasonic testing, eddy current testing. Micro structural characterization on two fuel pins, TH-5 and TH-2 from the cluster, has been done using optical microscopy, scanning electron microscopy, β-γ autoradiography and α-autoradiography techniques. This paper presents the salient observations of the examinations carried out on the irradiated fuel pins. 2. Fabrication details The experimental six pin cluster (AC-6) consisting of ThO 2-4% PuO 2 fuel pellets, encapsulated in a free standing Zircaloy-2 cladding, was fabricated at the Radio Metallurgy Division (RMD). The cluster consisted of short-length fuel pins of about 0.5m length and the sketch of a typical experimental MOX fuel pin is shown in Fig 1. Table 1 gives the details of the fabricated fuel pins. 3. Irradiation history The AC-6 experimental cluster was irradiated in the PWL of CIRUS research reactor. The thermal neutron flux in the loop was 5x10 13 n/cm 2 /sec and the temperature and pressure of the coolant in the loop was 240 o C and 105Kg/cm 2 respectively. Peak linear heat rating and peak burn-up in the fuel pins was 40 kw/m and 18.5 GWd/t respectively. 1

2 Table 1 Details of the Thoria based MOX fuel pins Cluster AC-6 Clad type Free standing Number of pins 6 (5 ThO 2-4% PuO Helium filled pin) PuO 2 enrichment 4% Pellet diameter ± 0.01 mm Pellet length 12.0 ± 1.0 mm Pellet density 92-94% TD Stack length 435 mm Cladding outer wall diameter 14.3 mm Cladding wall thickness 0.8 mm Cold plenum length 20 mm 4. Non-destructive testing 4.1. Visual examination Visual examination was carried out on the individual pins using a wall mounted periscope. All the fuel pins were found with deposit of loose white powder which may have come from the storage water pool containing aluminum cladded research reactor fuel assemblies. The fuel pins were cleaned by cotton damped with alcohol. No abnormality was visible on the surface Diameter measurement A diameter measuring set up was made to be suitable to keep the fuel pin straight on the platform. The diameter was read from the dial gauge display through the periscope. The diameter of the irradiated fuel pins was found to be within the manufacturing tolerances Leak testing Leak testing was carried out in the hot cell using liquid nitrogen and alcohol leak test. During the test each pin was immersed in liquid nitrogen for 5-7 minutes. Subsequently the fuel pin was transferred to get immersed in a tank filled with alcohol. There was no leak observed in the fuel pins Ultrasonic testing Fig. 1. A typical experimental MOX fuel pin Two 10 MHz line focused immersion probes were fitted at an angle in the probe carriage for detection of axial and circumferential defects. Multi-channel ultrasonic flaw detector was used for slow helical scan combining axial probe translation and rotation of fuel pin. Surface roughness signals were predominant making interpretation difficult. 2

3 Fig. 2. Loading of fuel pin to Ultrasonic scanner using master slave manipulator Fig. 3. Two channels of ultrasonic flaw detector showing surface signals 4.5. Eddy current testing Eddy current testing (ECT) was carried out on 4 fuel pins except pin TH-5, which could not pass through the eddy current testing coil. Defect signals were obtained from the cladding of pin TH-2 as shown in Figs 4 a and b. The fuel pin is pulled up slowly and signals are recorded by the computer. Plenum spring Defect signal Length of the fuel Fig. 4 a. ECT setup inside the hot cell Fig. 4 b. ECT signal from defect location of TH Gamma scanning Gamma spectroscopy and scanning using the HPGe detector and multi channel analyzer/scaler (MCA/MCS) were carried out on these elements inside the hot cell. Co-60 and Cs-137 sources were used for calibration. The spectra obtained from these elements revealed the presence of Cs-137, Cs- 134, Eu-154 and Tl-208. Fig. 5. Shows a typical gamma ray spectrum. Fig. 5. Gamma ray spectrum of (Th-4%Pu)O 2 fuel pin 3

4 4.5. Neutron radiography The fuel pin TH-4 suspected to be defective during eddy current testing was subjected to neutron radiography in the test reactor Circus. The bottom and top end plugs were found intact. Multi-axial cracks in the pellet were observed in the neutron radiograph shown in Fig. 6. The plenum spring was also observed to be free from any distortion. Insulation pellet Bottom plug Multi-axial pellet cracks Plenum spring Fig. 6. The neutron radiograph of fuel pin TH-4 shows no abnormality after irradiation. 5. Destructive testing 5.1. Fission gas analysis Fuel pins were punctured for the measurement of the amount of released fission gases. The volume of the released fission gas and the void volume in the fuel pin were measured to arrive at the pressure of the gas inside the fuel pin. A dual column gas chromatograph and a quadrupole mass spectrometer were used to analyze the chemical composition and isotopic composition respectively of the collected gases. Fission gas analysis was carried out on the fuel pins TH-2, TH-4 and TH-5. The fission gas pressure in the pin TH-4 and TH-5 was 4.4 and 3 atmosphere respectively and the gases were He, Xe and Kr. The fuel pin TH-2 did not show fission gas, indicating it to be a leaky pin which was also found defective by ECT Metallographic examination Metallographic examination of two fuel pins, TH-5 and TH-2 has been carried out to study the fuel restructuring, cladding oxidation and hydriding. β-γ autoradiography of the metallographic samples was carried out to study the distribution of the fission products (mainly Cs 137 ) across the cross section. α-autoradiography was carried out to analyze the distribution of plutonium in the fuel [2]. Radiographs are shown in Fig. 7a, 7b and 7c. Fig. 7a. Photo macrograph Fig. 7b. β-γ autoradiograph Fig. 7c. α-autoradiograph The white spots are the region of low activity, probably PuO 2 agglomerate created local high temperature and cesium migrated to other location. This observation is still under investigation. Radial 4

5 cracks were observed in the fuel. No columnar grain formation or grain growth was observed in the fuel. The β-γ autoradiographs revealed asymmetric distribution β-γ activity across the fuel cross section. Higher activity was observed in the cracks in the fuel. The α-autoradiograph of the fuel sections revealed a uniform Pu-activity along the cross-section. Uniform oxide layer was observed on the outer surface of the cladding, whereas a discontinuous oxide layer was noticed on the inner surface. Average oxide layer thickness on the outer and inner surface of the cladding was 1.3 µm and 0.9 µm respectively Clad metallography at defect location Metallographic samples were taken from fuel pin, TH-2, in which a defect location was identified during eddy current testing and no fission gas was obtained during fission gas measurement. Examination of the cladding revealed a massive hydride blister. Fig. 8.a. shows the photo macrograph of a section from the pin with a defect region in the cladding. View of the hydride blister at high magnification is shown in Fig. 8.b. Fig. 8a. Photo macrograph of the TH-2 fuel pin cross section. Fig. 8b. Massive hydride blister in the clad and subsequent cracking of clad Scanning electron microscopy (SEM) The cellulose acetate replica prepared from the fractured surface of the fuel from pin TH-5 was examined under SEM. The grain size was measured from the impression of the grains on the replicating tape. Fig. 9a. reveals the grain morphology observed in the fuel. The average grain size was found to be 14µm. A bimodal grain size distribution with grains up to 30µm in size and cluster of fine grains of 2-3µm size were also observed among the larger grains as, shown in Fig. 9b. (a) (b) Fig. 9a. SEM photograph showing grain morphology and 9b. bimodal grain size distribution 5

6 Examination of the fractured fuel surfaces revealed grains with as-fabricated pores at the grain edges and corners. At a few locations, very fine fission gas bubbles were observed on grain faces and there was evidence of interlinking of the fine bubbles as shown in Fig. 10a. A magnified image of the interlinked features present on the grain face is shown in Fig. 10b. Fig. 10a. presence of fine bubbles on the grain faces and 10b. Magnified image shows interlinking of fission gas bubbles in TH-5 6. Conclusions 1. Absence of visible grain growth or columnar grain formation was absent. 2. Radial cracks were present in the fuel pellet cross section. 3. Measured grain size and grain morphology was similar to the as-fabricated fuel with an average grain size of 14 µm. A few grains up to 30 µm in size and at some locations were observed. Clusters of fine grains of 2-3 µm were also observed. Bimodal grain size distribution occurred in some regions of the fuel. 4. Reduced porosity was seen in the central portion of the fuel. 5. Submicron size fission gas bubbles on the fuel grain surfaces in the central region and inter linkage was also observed. REFERENCES [1] JOURNAL OF NUCLEAR MATERIALS 383 (3008) , Utilization of Thorium in Reactors. [2] 11 th INTERNATIONAL CONFERENCE ON CANDU FUEL, CANADA, OCTOBER 17-20, Post-Irradiation Examination of Candu Fuel Bundles Fuelled with (Th, Pu)O 2. 6