JAERI-Research—-95-087 JAERI-Research JP9602233 95-087
BEHAVIOR OF IRRADIATED PWR FUEL UNDER A SIMULATED RIA CONDITION [RESULTS OF NSRR TEST MH-3]
December 1995
Hideo SASAJIMA, Toyoshi FUKETA, Yukihide MORI Kiyomi ISHIJIMA, Shinsho KOBAYASHI, Takeshi YAMAHARA Tomohide SUKEGAWA and Tadaharu ITO
Japan Atomic Energy Research Institute 1 7 H° 1 t (T319-11 %4«A
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This report is issued irregularly. Inquiries about availability of the reports should be addressed to Information Division, Department of Technical Information, Japan Atomic Energy Research Institute, Tokai- mura, Naka-gun, Ibaraki-ken 319-11, Japan.
©Japan Atomic Energy Research Institute, 1995
H * m -i- JAERI-Research 95-087
Behavior of Irradiated PWR Fuel Under a Simulated RIA Condition [Results of NSRR Test MH-3]
Hideo SASAJIMA, Toyoshi FUKETA, Yukihide MORI, Kiyomi ISHIJIMA Shinsho KOBAYASHI, Takeshi YAMAHARA\ Tomohide SUKEGAWA' and Tadaharu IT0f
Department of Reactor Safety Research Tokai Research Establishment Japan Atomic Energy Research Institute Tokai-mura, Naka-gun, Ibaraki-ken
(Received November 8, 1995)
Results from the power burst experiment, Test MH-3, conducted in the Nuclear Safety Research Reactor (NSRR) are presented. A fuel rod was irradiated with a fuel burnup up to 38.9 MWd/kgU in the Mihama unit #2 of the Kansai Electric Power Co., Inc. The Test MH-3 was the third and final experiment in a reactivity initiated accident (RIA) test series with the MH fuel rod. Data concerning test method, pre-pulse examination, pulse irradiation, transient records and post-pulse fuel examination are described, and discussions of the results are presented. A test fuel rod is a short-sized 14x14 pressurized water reactor (PWR) type rod, which is refabricated from a full size commercial fuel rod. A double container-type capsule contains the instrumented test fuel rod with stagnant water cooling condition at atmospheric pressure and ambient temperature. The test fuel rod was subjected to the pulse irradiation resulting in a deposited energy of 87 cal/g*fuel and a peak fuel enthalpy of 67 cal/g-fuel. Behavior of the test fuel rod was assessed from pre- and post-pulse examinations and transient records during the pulse irradiation. Cladding surface temperature increased about 200°C. The maximum cladding deformation was 1.6% and the test fuel rod did not fail. Estimated fission gas release during the pulse irradiation is 3.8% in Kr, and 2.3% in Xe, respectively. Through the detail fuel examination, information concerning microstructural change in the fuel pellets was also obtained.
Keyword: NSRR, RIA, Irradiated Fuel, PWR, Pulse Irradiation, Fuel Burnup, PIE
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Contents
1. Introduction 1 2. NSRR Facility 2 3. Irradiation in the Mihama Unit 82 Reactor 3 3. 1 Irradiation Conditions 4 3. 2 Post Irradiation Examinations of the Mother Fuel Rod 4 4. Refabrication of the Test Fuel Rod 11 5. Test Conditions 12 5. 1 Test Procedure • 12 5. 2 Test Fuel Rod 12 5. 3 Pre-pulse the Test Fuel Rod Examinations 13 5.4 Pulse Irradiation in the NSRR 14 5.4.1 Test Capsule 14 5.4. 2 Instrumentation 15 5.4.3 Irradiation Conditions 15 6. Transient Behavior during the Pulse Irradiation 16 7. Post-pulse Examinations of the Test Fuel Rod 16 7. 1 Non-destructive Tests 17 7.2 Destructive Tests 19 8. Summary and Discussion 22 9. Conclusions 26 Acknowledgment 27 References • 28 JAERI-Research 95-087
1. it Cadiz 1 2. NSRROU® 2 3. m^2^mr-(DmM 3 3. 1 mM&H¥ 4 3.2 -?-tf-ffi® IV JAERI-Research 95-087 1. Introduction To provide a data base for the regulatory guide of light water reactors (LWRs), the behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of Japan Atomic Energy Research Institute (JAERI). Numerous experiments using pulse irradiation capability of the NSRR have been performed to evaluate the thresholds(1), modes(2), and consequcnces(3) of fuel rod failure in terms of the fuel enthalpy, the coolant conditions, and the fuel design. All of the experimental data used for current safety evaluation guideline for reactivity initiated events in light water cooled power plants(4) established by the Nuclear Safety Commission were limited to those derived from the experiments with fresh fuel rods. In time past, the tests with pre-irradiated fuel rods were conducted in the SPERT-CDC(S>6) and PBF(7> 8>9) projects, however, limited and being insufficient for clearly understood the influences of fuel burnup on the fuel behavior under RIA conditions. The current Japanese guideline introduces the peak fuel enthalpy of 85 cal/g*fuel adopted from the SPERT-CDC data as a provisional failure threshold of pre-irradiated fuel rods under RIA conditions, and this failure threshold is considerably lower than that of fresh fuel rods. The failure threshold for irradiated fuel rods was not satisfactorily established and the failure mechanism of irradiated fuel rods during RIA remains undetermined, because limited number of irradiated fuel rods were tested. Besides the requirements for the regulation, the burnup effect becomes one primary concern in the field of fuel behavior study since demand of extended fuel burnup is widely recognized. In these conditions, the NSRR phase II program with prc-irradiated fuels as test samples was started in July 1989. The objectives of this program arc; (1) to determine the failure thresholds of the pre-irradiated fuel rods with the parameters including the fuel burnup, the initial rod power, etc., (2) to clarify the failure modes and the failure mechanisms of the prc-irradiated fuels, and (3) to clarify the consequences of the failure of the pre-irradiated fuel rods. Results obtained from phase II program will reinforce data base for safety assessment of light water reactors. The test fuel rods used in this program are; (1) Short-sized fuel rods refabricated from the commercial-sized fuel rods irradiated in the power reactors, pressurized water reactors (PWRs)(10> n) and boiling water reactors (BWRs)(12>13), and JAERI- Research 95-087 (2) Short-sized fuel rods pre-irradiated in the Japan Materials Testing Reactor (JMTR)(H15'16'17). The test fuel is contained in a double container capsule filled with water of the ambient temperature at an atmospheric coolant pressure, and subjected to a pulse irradiation simulating a prompt power surge in the RIA. Principal experimental parameters arc a fuel bumup and a peak fuel enthalpy during the pulse irradiation. Fuel behavior is studied through data found from the on-line instrumentation and post irradiation examinations (PIEs) of the fuel. In this report, the results obtained in the Test MH-3 arc described and discussed. Test fuel was obtained from the fuel rod irradiated in the Mihama unit #2 reactor (from now on the mother fuel rod). The mother fuel was irradiated to the radial average burnup of 38.9 MWd/kgU at a linear heat generation rate of 19.8 kW/m (from now on the prc-irradiated conditions). Fractional fission gas release was 0.15% in Kr and 0.16% in Xc under the reactor operations, respectively. The pulse irradiation of the Test MH-3 was performed at the peak fuel enthalpy of 67 cal/g-fuel on October 31,1990, as the third NSRR power bursts experiment with irradiated in the Mihama unit #2 reactor following the Tests MH-1(18) and MH-2(19). Maximum cladding surface temperatures reached about 200° C, and additional fission gas release was 3.8% in Kr and 2.3% in Xc at the peak fuel enthalpy of 67 cal/g*fucl during the pulse irradiation, respectively. After the pulse irradiation, the mctallographical results showed that microcracks developed on the fuel periphery along the grain boundary. The additional fission gas release during the pulse irradiation is controlled by the formation of intergranular microcracks on the fuel pellet periphery. 2. NSRR Facility The NSRR is a modified TRIGA-Annular Core Pulse Reactor (ACPR) using uranium- zirconium hydride (U-ZrH, 6) fuel-moderator elements. The reactor has a dry experimental cavity (the experimental tube) of 22 cm in inner diameter, penetrating the core center. The test fuel rod contained in an experimental capsule is installed in the experimental tube and exposed to a high pulsed neutron flux simulating RIA conditions. As shown in Fig. 1, the core structure is locating at the bottom of 9 m deep open pool, and is cooled by natural circulation of the pool water. The NSRR was designed exclusively to conduct power transient experiments and attained its first criticality in May 1975. Figure 2 illustrates the rod arrangement of the NSRR. The core contains 149 driver fuel-moderator elements, 6 fuel follower regulating rods and 2 fuel follower - 2 - JAERI-Research 95-087 safety rods, making 157 driver fuel elements in total. In addition, 3 transient rods with air filled followers are used to insert reactivity for pulsing operation. The regulating rods and the safety rods are driven electrically, while the transient rods arc driven pneumatically. These rods arc arranged to form a hexagonal pattern surrounding the experimental tube. An active zone of the fuel moderator clement is 38.1 cm long and 3.56 cm in diameter in which zirconium hydride moderator (88%) is homogeneously combined with enriched uranium fuel (12%). The uranium is enriched to 20% in ^U and the H/Zr atom ratio of the zirconium hydride moderator is 1.6. The pulsing power escalation is controlled by spectrum hardening caused by the moderator temperature increase, besides the Doppler effect. The characteristics of the NSRR arc listed in Table 1. The detailed description of the NSRR was provided by Saito ct al.(2n). After the modification of the operating system in fiscal year 1988, the NSRR has been operated in four modes as shown in Table 2. Steady state operation of power of up to 300 kW is controlled by adjusting a bank of regulating rod position. High power operations of up to 10 MW in the Shaped Pulse and the Combined Pulse modes arc achieved by quick operations of the regulating rods at a speed of up to 75 mm/s. The reactor energy release is limited within 110 MW in the two modes. Pulsing from zero and rated powers is realized by quick withdrawal of the 3 transient rods in the Natural Pulse mode and Combined Pulse mode, respectively. The maximum reactivity insertion of $4.7 from zero power is allowed to produce peak reactor power of 23,000 MW and reactor energy release of 130 MW for the pulsing operation. The Test MH-3 was performed in the Natural Pulse operation mode from zero power. 3. Irradiation in the Mihama Unit #2 Reactor Figure 3 illustrates the design of the mother fuel rod. The fuel stack of the mother fuel rod is composed of 239 UO2 pellets initially enriched to 2.6% in ^U, and is 3642 mm in height. Total length is 3856 mm and plenum length is 179 mm. The fuel pellets arc sheathed with the zircaloy-4 thin tube of 10.72 mm in outer diameter and 0.62 mm in thickness. The initial radial gap width is 0.095 mm. The mother fuel rod was filled with helium of about 3.2 MPa. The initial specifications of the mother fuel rod arc listed in Table 3. o JAERI-Research 95-087 3.1 Irradiation Conditions The mother fuel rod of the MH test series has been irradiated in the Mihama unit #2 reactor of Kansai Electric Power Co. during 4 cycles from 1978 to 1983, to the radial average burnup of 38.9 MWd/kgU at the linear heat generation rate of 19.8 kW/m (13.9 kW/m in 1st., 24.3 kW/m in 2nd., 21.7 k\V/m in 3rd. and 19.3 kW/m in 4th. cycles). The fractional fission gas release during the pre-irradiation was 0.15% in Kr and 0.16% in Xc, respectively. The fuel bumup of the test fuel rod during the prc-irradiation was evaluated from the fractional amounts of isotopic neodymium and plutonium per uranium ratio measured with the mass spectroscope for the solution using the fuel pellet of the Test MH-2. Because the fuel segments for the Tests MH-2 and MH-3 was taken where the gamma profile was flat, the fuel burnup of the Test MH- 3 segment is identical to one Test MH-2. The pre-irradiation conditions are summarized in Table 4. 3.2 Post Irradiation Examinations of the Mother Fuel Rod The post irradiation examinations of the mother fuel rod were performed to investigate the fuel conditions before segmented to the test fuel rods for the pulse irradiations. These post irradiation examinations consist of nondestructive tests and destructive tests. The nondestructive test includes (1) visual observation, (2) dimensional measurement, (3) X-ray radiography and (4) gamma-ray scanning. The destructive test includes (1) fission gas release, (2) optical microscopy, (3) scanning electron microscopy (SEM) and electron probe microanalysis (EPMA) test, (4) micro gamma-ray scanning on the fuel pellets, (5) hydrogen analysis of the cladding, (6) measurement of oxidation thickness for the cladding, (7) measurement of the cladding tube hardness, (8) X-ray diffraction on the fuel pellets, (9) autoradiography, (10) photo-image analysis on the fuel pellet and (11) out gas analysis (OGA) on the fuel pellets. 3.2.1 Non-destructive Tests (1) Visual Observation During the visual inspection of the mother fuel rod, discoloration, absorption of crud, scratches, speckles and grid marks on the cladding outer surface was observed. Discoloration can be caused by oxidation that was particularly emphasized on diffusion in both the oxides in - 4 - JAERI-Research 95-087 coolant and the cladding substrate. However, significant defects of the cladding tube were not formed during the prc-irradiation. (2) Dimensional Measurement The dimensions of the mother fuel rod were measured with the profilemeters. The examinations comprise diameter and length measurements. The diameters were measured along two ridge lines in 0=zero and 0=90 degrees over the entire axial length of the rod without top and bottom regions for the setting on the profilemeter. The direction named zero degree was defined arbitrarily. The pitch of the trace is 1 mm, and the accuracies arc ±0.01 mm for the diameter and ±0.5 mm for the length. Figure 4 shows the axial profile of the mother fuel rod diameter. The diameter ranged from 10.637 mm to 10.750 mm in 0=0 degree and from 10.635 mm to 10.735 mm in 0=90 degree, and the entire axial length was 3875.2 mm. The data of dimensional measurements indicated that creep down of the cladding was slightly produced, and no deformation was caused by pellet cladding mechanical interaction (PCMI) during the prc- irradiation. (3) X-ray Radiography X-ray radiographs were taken from the two directions. As the results of the X-ray radiographs, the active fuel length was 3658 mm in 0=0 degree and 3657 mm in 0=90 degree, and the plenum length was 178 mm in 0=0 degree and 179 mm in 0=90 degree. Although the axial gap of about 0.2 mm between the 18th and 19th pellets from the bottom, and fewer numbers of small gaps between the pellets were observed, abnormalities e.g. fragment of fuel pellets were not observed after completion of the pre-irradiation. (4) Gamma-ray Scanning Gamma-ray scanning of the gross gamma and each fission product for the fuel rod gives information on the axial distribution of fuel burnup or migration of fission products. Axial distributions of gamma-ray intensities for gross gamma and ccsium-137 were continuously measured at forward speed of about 0.42mm/s along the fuel rod as shown in Fig. 5. Cobalt-60, rhodium-106, cesium-134, cesium-137 and europium-154 were step-wisely measured at pitch of about 0.017 mm/s as shown in Figs. 6 though 10, respectively. The relatively flat gross 5 JAERI-Research 95-087 gamma was observed between the spacer grids, and ccsium-134 and cesium-137 accumulated in the axial center of the pellets. Gamma spectrometry was conducted for 8000s at the center of the fuel active region and at the axial elevation of 2011 mm from the top end plug as shown in Fig. 11. Signals correspond principally to long half life of fission products. Gamma spcctromctry was conducted also at the gas plenum and at the axial elevation of 90 mm as shown in Fig. 12. Signal corresponding to cobalt-60 was detected from the spectrum of the activated stainless steal spring, and the signals corresponding to krypton and xenon were not detected. 3.2.2 Destruaive Tests As shown in F5g. 13, two round slices and one vertical division were taken from the axial position z = 1322.0 to 1452.0 mm of the mother fuel rod for metallographical examinations and the other examinations after the rod puncture test. One round slice, named MMC1, was subjected to optical microscopy, micro gamma-ray scanning on fuel pellet, cladding tube hardness measurement and pellet photo-image analysis. The other round slice, named MMC2, was subjected to optical microscopy, SEM, EPMA, p/y and a autoradiography and X-ray diffraction. Vertical division, named MML, was subjected to optical microscopy, cladding oxidation thickness measurement and pellet photo-image analysis. Hydrogen concentration on the cladding was measured with the small test pieces of the cladding. (1) Fission Gas Release The mother fuel rod was punctured with fission gas collection device at the plenum. Free volume in the fuel rod, amount of released gases and internal pressure were measured. Volume ratios of the released gases and abundances of isotopes were measured with the mass spectrograph. Table 5 shows test results of rod puncture and gas analysis. Fission gas inventory in the fuel rod was calculated with ORIGEN-2(21) code as shown in Table 6. After the completion of the prc-irradiation, 0.15% Kr release and 0.16% Xe release were estimated with the rod puncture, gas analysis and ORIGEN-2 code analysis. (2) Metallography Microstructural evaluation and radio-nuclidc distribution across the radius of a fuel pellet depend primarily on irradiation history and resulting fuel temperatures. An important - 6 - JAERI-Research 95-087 microstructural feature is the radius at which grain boundary bubble precipitation occurs. Data on the fuel behavior during the pre-irradiation were obtained from metallographies and studies of fuel rod axial and vertical cross-sections. As shown in Fig. 13, two round slices and one vertical division were taken from the mother fuel rod as samples for metallographical examinations. The specimens MMCl and MML were subjected to optical microscopy. A specimen for SEM and EPMA was taken from the specimen MMC2. Photograph 1 shows macroscopic view of the samples as polished. a) Optical Microscopy To obseive fuel pellet radial and axial cross-sections and pellet/pellet boundaries, the specimens MMCl and MML were subjected to optical microscopy. Enlarged photographs (200x, 400x and 800x magnifications) of the specimen MML were taken in pellet center, intermediate and outer edge regions with polished and etched surfaces, respectively, and arc shown in Photo. 2 with macroscopic view (5x magnification). Enlarged photographs of the specimen MMCl were taken in pellet center, intermediate and outer edge regions with polished and etched surfaces, respectively, and are shown in Photo. 3 with macroscopic view (5x magnification). The macroscopic photographs of the specimens MML and MMCl show the large cracks generated during the pre-irradiation, and no evidence of fuel melting and fuel shattering. The outer edge of the fuel pellet is characterized by the loss of optically definable grain structure. Pores are obviously present on grain boundaries and triple points. Pore migration is not extensive and a network of intergranular porosity is not observed in all regions from these macroscopic photographs. b) SEM and EPMA on Fuel Pellet A fuel piece obtained from the specimen MMC2 was examined with SEM to evaluate fuel miewstructure, and the EPMA measurements were made to evaluate the distribution of retained cesium and plutonium. The SEM pictures of the center, intermediate, peripheral and outer edge regions for the fuel pellet at large magnifications (lOOOx, 3000x magnifications) arc shown in Photos. 4. Grain growth and concurrent grain boundary sweeping of fission gas bubbles have been postulated as mechanisms associated with fission gas release. For the SEM examined cross sections of fuel pellet, the radial grain growth and grain boundary separation were not observed in all regions, and development of a porous rim structure was not observed in the pellet - 7 - JAERI-Research 95-087 peripheral or outer edge region. It shows that fission gas released by diffusion principally from the fuel pellet during the prc-irradiation. SEM pictures show that pores were obviously present on grain boundaries and triple points and increased in the pellet outer region. Though many pores were seen in the pellet outer region, there is no evidence of the pores interconnecting. The radial cracks in the fuel pellet generated through grain boundaries as shown in the SEM images of Photo. 5. Photograph 6 shows the SEM image at location A. Distribution measured in line analysis and X-ray images of oxygen, ncodymium, ruthenium and cesium arc shown in Photo. 7, anomaly and characteristic heterogeneity were not found. Point clement analysis using energy dispersion method of EPM A was also conducted in the center, intermediate and peripheral regions, and the X-ray spectrums showing characteristic peaks of uranium and cesium in the three locations arc shown in Fig. 14. (3) Micro Gamma-ray Scanning on Fuel Pellet Micro gamma ray scanning was performed at forward speed of 1.5 mm/min along two radial directions of the specimen MMC1. The gross-gamma ray and gamma ray from cesium- 137 were measured as shown in Fig. 15. The gross-gamma and ccsium-137 intentions increased in the peripheral region, though it is almost flat in the central region. According to buildup plutonium, the local fuel burnup also becomes much higher in the peripheral region than in the central region because cesium-137 contributes to the fuel bumup. It is considered that no cesium relocation occurs along the pellet radius up to about 40 MWd/kgU in the normal operating conditions. (4) Oxidation and Hydrogen Absorption of the Cladding Tube A radially oriented hydride distribution in the cladding tube of the specimen MMC1 with the fuel bumup of 38.9 MWd/kgU can be seen in Photo. 8. Photograph 9 shows the characteristic higher hydrogen concentration at the cladding outer surface and lower hydrogen concentration with some hydrides precipitated radially at the cladding inner surface obtained from the specimen MML. The migration of the hydride precipitated towardly to the outer surface of the cladding. It could be a sufficient condition for the enhanced corrosion. Measured from the small test pieces of the cladding, the hydrogen concentration was 41.4 ppm in entirely over the cladding wall thickness of about 0.6 mm. Photographs. 10 and 11 show the oxidation layer of the specimens MMC1 and MML, respectively. Oxygen diffused into the cladding from the external metal- JAERI-Research 95-087 water reaction or from the internal cladding and fuel contact becomes significant at high cladding temperatures. Oxygen uptake, and zirconium hydride formation, generally strengthens the cladding, but greatly reduces the ductility. (5) Measurement of the Cladding Tube Hardness Measurements of micro-hardness were taken across the cladding of the specimen MMC1 in an attempt to develop a quantitative estimation of the cladding ductility using the micro vicars hardness test device. The results of these measurements are shown graphically in Fig. 16. Hardness values of ranged from 269 Hv to 277 Hv were measured across the cladding wall thickness and were about 1.3 times as higher as that of unirradiatcd cladding. Increases of the cladding hardness did not appear toward to the cladding outer surface. It could be indicated that the effect of the hydride on the cladding hardness was not significantly at the cladding outside in this test. (6) X-ray Diffraction The defects of the lattice parameter for UO2 fuel are directly caused by the accumulation of radiation damage and change of chemical state i.e. O/M ratio and dissolution of solid fission products, in the lattice. The lattice parameter at the center of the specimen MMC2 was measured by X-ray diffractometer with copper Ka radiation that has beam current of 30 raA at 40 kV. The diameter of the X-ray beam was 1.0 mm for the specimen. The lattice parameter was obtained for the fourteen diffraction lines between 5° and 130° in 20. The scanning speed of the goniometer was 2°/min. Figure 17 shows results of X-ray diffraction measurements for the mother fuel pellet. The lattice parameter ranged from 546.3 to 547.8 pm, listed in Table 7, at the fuel bumup of 39.8 MWd/kgU were totally higher than that of unirradiatcd fuel. This lattice dilation was principally due to the accumulation of fission gas bubbles and radiation defects. The porosity evidently contains part of the fission gas lost from the UO2 lattice in the region where the microstructure changes take place. (7) Autoradiography Beta and gamma-autoradiography and a-autoradiography were performed using the specimen MMC2 as shown in Photo. 12. Radial power skew of Py emission on the pellet was - 9 - JAERI- Research 95-087 not clearly observed. Higher concentration of oc emission was observed in a thin layer of the pellet outer edge. (8) Grain Size of Fuel Pellet Grain growth and concurrent grain-boundary sweeping of fission gas bubbles have been postulated as mechanisms associated with fission gas release. Fuel restructuring in the test is characterized by limited cquiaxed grain growth. Grain sizes in the UO2 fuel structures were determined from micrographs of the specimens MMCl and MML with the photo-image analyzer. The photograph of magnification 800x was selected for the grain size measurements. The grain sizes were measured at three radial regions, the center, intermediate and periphery. Intermediate regions of the specimen MMCl were measured in two directions. The number of grain boundaries intercepting a certain length line was automatically counted and the average distance between intercepts was then calculated automatically and used for the nominal grain diameter. Objects smaller than 1.5 /im were regarded as noise and rejected. The size distributions of the specimens MMCl and MML arc shown graphically in Figs. 18 through 24, respectively. The averaged grain size of the specimen MMCl was 7.3 //m, 6.6 fxm (average of 6.3 /xm and 6.9 /xm) and 7.0 /xm at center, intermediate and periphery regions, respectively. The averaged grain size of the specimen MLC1 was 5.8 //m, 5.1 /xm and 5.5 fxm at center, intermediate and periphery regions, respectively. The radial distribution of the grain size was not significantly. The rim restructuring at the fuel periphery did not appear though this grain size measurement. (9) Out Gas Analysis of Fuel Pellet Annealing tests on the fuel pellet specimens were performed to determine the fission gas release rates as functions of temperature and time with an out gas analyzer (OGA)(22) in the RFEF. The specimens in graphite containers were heated in an induction furnace under vacuum condition at temperatures of 550 °C to 2000 °C. The furnace temperature was measured with two optical pyrometers. The released gas from the fuel specimen was collected in the reservoir tank, and then it was measured with the mass spectrometer. The four pieces of UO2 pellets, MHT1, MHT2, MHT3 and MHT4, were taken from the mother fuel rod in the axially ranged between 1280.0 mm and 1304.5 mm from the bottom as shown in Fig. 13. These specimens had almost flat axial distribution of gross gamma-ray intensities. The test conditions of annealing for the four specimens are shown in Figs. 25 though 28, respectively. Each fuel specimen was -10- JAERI-Research 95-087 heated up to several temperatures by steps with the heating rate of about 7 °C/s under vacuum conditions. The total inventories of the fission gases for each fuel specimens were calculated with the ORIGEN-2 code. in the test with the specimen MHT1, fractional fission gas releases were less than 0.25% at each maximum temperature of 550 °C, 750 °C and 1000 °C as shown in Fig. 29. The specimen MHT2 was heated up to several temperatures in the ranged from 750 °C to 2000°C as shown in Fig. 26, and it was continuously heated for about 10 minutes at each temperature. The fractional fission gas releases were not less than 3% at 1400 °C, and were more than 10% over 1600 °C as shown in Fig. 30. To investigate the effect of continuous time, the specimen MHT4 was heated up to 1400 °C and it was continuously heated for each 10, 20 and 45 minutes. The fractional fission gas releases were less than 3% as shown in Fig. 32 and were not dependent on continuous times in this test. Simulate to the power ramp test was performed with the specimen MHT3 in the temperature range from 800°C to 1400°C as shown in Fig. 27. The fractional fission gas release was less than 2%, and the burst release did not occur during the test as shown in Fig. 31. In the Tests MHT1, MHT3 and MHT4, the higher fractional fission gas release appears at the early stage of the tests. 4. Refabrication of the Test Fuel Rod As shown in Fig. 33, the four test fuel rods were rcfabricatcd from the mother fuel rod for the NSRR experiments. The MH-3 test fuel rod was taken from the mother fuel rod in the axial range of 1832 mm to 1980 mm from the bottom. After marked to cutting position of direction and identification of top and bottom on the cladding outer surface, the mother fuel rod was cut by the dry severance. And then the fuel pellets at both ends were removed from the test fuel regions, and the top and bottom end fittings were welded at both ends of the test fuel region. The conditions of the welding arc summarized in Table 8. The test fuel rods were filled with pure helium and then tips of the top end fittings were welded by the seal welder. The conditions and results of filling gas and seal welding for the test fuel rods arc summarized in Table 9. Helium leak tests were performed to check possible leakage in all connections of the test fuel rods by the vacuum-prcssurization method. Consequently, any leakages were not found. The conditions and results of helium leak tests for the test fuel rods are summarized in Table 10. -11- JAERI- Research 95-087 5. Test Conditions 5.1 Test Procedure The experimental procedure of the NSRR experiments with fuel rods irradiated commercial reactors is illustrated in Fig. 34. The test fuel rod was refabricated from the mother fuel rod at the RFEF in JAERI Tokai. After the refabrication and the pre-pulse fuel examinations in the RFEF, the test fuel rod was transferred to the NSRR and instrumented in an iron hot cell with instrumentation devices. The instrumented test fuel rod, attached to a fuel supporting frame connecting to a capsule flange shown in Photo. 13, was assembled into the test capsule in a hot cave locating next to the hot cell. The inner capsule containing the test fuel rod was filled with water at ambient temperature and closed leak tight with cover gas of helium at the atmospheric pressure. Tightness for fission product leakage was examined by helium leak tests of the inner and the outer test capsules. The test capsule was transferred into the experimental cavity thought the offset tube with a loading device equipped with radiation shields and instrumentation cable connections. The test fuel rod contained in the capsule and was pulse irradiated. Using the same device, the capsule is transferred from the cavity to the hot cave after the pulse irradiation. The outer capsule is disassembled in the cave after the confirmation of the shield of the inner capsule. Then, the inner capsule containing the test fuel rod was transported from the NSRR to the Research Hot Laboratory in JAERI Tokai, and is disassembled. The test fuel rod was subjected to the extensive fuel examinations. 5.2 Test Fuel Rod The test fuel contains eight UO2 pellets initially enriched to 2.6% in ^U that is composed part of the fuel stack of about 122 mm in height. Figure 35 illustrates the design of the test fuel rod used in the Test MH-3. The specification of the test fuel rod is identical to that of the 14x14 PWR fuel rod except length, and is almost identical to that of the standard test fuel rod of the NSRR experiment. At both ends of the fuel stack, hafnium disks are located to avoid undesirable axial power peaking during the pulse irradiation and alumina pellets arc located to avoid the thermal effect on the fuel pellets during the welding. On the top of the fuel stack, a movement marker and an iron core arc placed to measure the fuel stack elongation. These items are sheathed with the zircaloy-4 thin tube. A bottom end fitting contains a pressure sensor for the -12- JAERI-Research 95-087 measurement of rod internal pressure. The test fuel rod was filled with pure helium of about 4.7 MPa (at Zero°C). The use of pure helium instead of realistic mixture gas may not significantly influence the fuel behavior because of the very low fission gas release before completion of the prc-irradiation. The initial radial gap width is 0.095 mm. Table 11 summarizes major characteristics of the test fuel rod. 5.3 Pre-pulse the Test Fuel Rod Examinations Preceding to the pulse irradiation experiment, the MH-3 test fuel rod was subjected to pre- pulse nondestructive examinations. The examinations include (1) visual observation. (2) X-ray radiography, (3) dimensional measurement and (4) the eddy current test. Coordinates in the test fuel rod are defined as shown in Fig. 36. (1) Visual Observation and X-ray Radiography The test fuel rod was visually inspected and photographed before the pulse irradiation. The photographs were taken from the one direction, namely from 0 degrees. The direction named 0 degrees was defined arbitrarily. Photograph 13 gives the appearance of the fuel rod before the pulse irradiation. Discoloration of the cladding surface along the active fuel region was noted for the test fuel rod through visual observation. It can be thought that oxidation with particular emphasis on diffusion in both the oxides in coolant and the cladding substrate. The cladding defects are not observed in this photograph. X-ray radiographs were also taken to confirm the existence of abnormalities i.e. fragment of fuel pellets from the two directions. The test fuel rod was visually observed and was X-ray radiated, and by that the test fuel rod did not has any significant abnormality. (2) Dimensional Measurement The dimensions of the test fuel rod were measured with the profilemetcrs. The examinations comprised diameter and length measurements. The diameter was measured along two ridge-lines in 6=zcro and 0=90 degrees over the entire axial length of the rod. The pitch of the trace was 1 mm, and the accuracies were ±0.005 mm for the diameter and ±0.5 mm for the length. Figure 37 shows the axial profile of the test fuel rod diameter after the refabrication. The diameter in the fuel active region ranged 10.643 mm to 10.690 mm in 0=0 degree and -13- JAERI-Research 95-087 10.643 mm to 10.676 mm in 0=90 degree, and the length was 298.7 mm. The data of dimensional measurements indicated that there was no deformation caused by PCMI during the prc-irradiation and was no effect of segmentation and welding. (3) Eddy Current Test As shown in Fig. 38, eddy current tests for the test fuel rod were performed with frequencies of 128 kHz (XI and Yl) and 1024 kHz (X2 and Y2) in the horizontal and vertical directions to investigate the existence of cladding defects. Phase angles for two measurements were 315° and 250°, respectively. The data of X3 and Y3 combined with XI and X2, and Yl and Y2, respectively. Although the test fuel rod had a few signals which could be caused by cruds absorbed on the outer surface, notable defects of the cladding tube were not formed during the irradiation in the Mihama reactor and during the rcfabrication. Figure 39 shows the data of the eddy current test for a reference tube which has artificial defects. Signals of artificial defects, A, B, C, D, E and F indicates in Table 12. 5.4 Pulse Irradiation in the NSRR 5.4.1 Test Capsule The test capsule used in the Test MH-3 is the double container-type capsule (type X-II atmospheric pressure capsule) for the experiment with pre-irradiatcd fuel. The outer capsule is a scaled container of 200 mm in the maximum outer diameter and 1250 mm in height. The inner capsule is a scaled pressure vessel of 72 mm in inner diameter and 680 mm in height. The schematic configuration of the outer and inner capsules is shown in Fig. 40. The type X-II capsule was specially designed for the test with fuel rod pre-irradiatcd in the commercial reactor and can be used in the test where energy deposition is less than 200 cal/g • UO2. The capsule contains an instrumented test fuel, is filled with stagnant water of ambient temperature and atmospheric pressure, and is subjected to the pulse irradiation simulating the RIA transient in the NSRR. -14- JAERI-Research 95-087 5.4.2 Instrumentation As shown in Fig. 41, the test fuel rod was instrumented with one thermocouple (type R, 0.2 mmcj)) to measure cladding surface temperature, the axial elongation sensors for the fuel pellet stack and. cladding, the movement marker for the axial elongation of the fuel stack, and the pressure sensor to measure rod internal pressure during the pulse irradiation. The thermocouple is attached to the cladding surface at 118 mm from the bottom of the fuel rod. The capsule was instrumented with one thermocouple (type K) to measure the coolant water temperature, and the pressure sensor and the strain gauge to detect the generation of pressure pulse inside the inner capsule due to fuel failure. Condition of data acquisition is listed in Table 13. 5.4.3 Irradiation Conditions The test fuel rod contained in the type X-II capsule was irradiated by the natural pulse mode with a reactivity insertion of $4.45 nominal in the NSRR at 2:45p.m., October 31, 1990. The primary irradiation conditions of the Test MH-3 arc summarized in Table 14 with these of the Tests MH-1 and MH-2. The transient records of NSRR reactor power measured with two micro fission chambers are shown in Fig. 42. The reactor power reached approximately 18 GW at maximum, and the pulse width at the half maximum power was 8.75 ms. Figures 43 shows the histories of core energy release (integrated reactor power) measured with micro fission chamber #1 and #2. The core energy release reached about 100 MJ during the pulse irradiation. During the pulse irradiation, the energy of 87 cal/g-fucl was deposited to the test fuel, and the fuel enthalpy reached 67 cal/g"fuel at maximum. To estimating energy deposition in the test fuel rod, a coupling factor of the test fuel rod power a*nd reactor power, is utilized in the NSRR project, which is the number of fissions in the unit mass of fuel generated by the pulse irradiation of unit energy release of the NSRR. As the coupling factor, numbers of fission in the test fuel per unit reactor energy release was estimated in the Test MH-2. The energy deposition of the Test MH- 2 was evaluated by chemical separation of barium-140 from mixed fission products in the sample solution and by gamma-ray measurement of barium-140. The coupling factor evaluated in the Test MH-2, which was performed under closer experimental condition to the Test MH-3 is used for estimating the energy deposition in the Test MH-3. -15- JAERI-Research 95-087 6. Transient Behavior during the Pulse Irradiation (1) Cladding Surface and Coolant Temperatures Figure 44 shows the transient history of the cladding surface temperature measured by thermocouple. The cladding surface temperature reached about 200 ° C at maximum. This shows that DNB (Departure from Nucleate Boiling) occurred on the cladding surface during the pulse irradiation. Figure 45 shows the transient history of the coolant water temperature near the top of the test fuel rod. The increase of the coolant water temperature corresponding to the increase of the cladding surface temperature. The coolant water temperature reached about 37 ° C at maximum at 7 ms after the time of peak power. During a time period from the peak power to 1 s, wavy nature was observed in transient histories of the coolant water temperature oscillated from 25 ° C to 44 ° C. (2) Fuel Stack and Cladding Elongation, and Rod Internal Pressure Figure 46 shows the axial elongation of the fuel stack and cladding tube during the pulse irradiation. The pellet stack elongation initiated at the pulse peak and increased to the maximum of 1.27 mm at 238 ms after the peak. The cladding also started to elongate at the pulse and the elongation reached 1.51 mm at 204 ms after the peak. The rod internal pressure increased from 5.0 MPa to the maximum of 5.2 MPa after 227 ms from the peak as shown in Fig. 47. (3) Capsule Internal Pressure and Strain The capsule internal pressure sensor and strain gauge attached to the bottom of the inner capsule flange did not give any notable signals. The peak signals obtained in the capsule internal pressure sensor at the times of 196 ms, 204 ms and 296 ms after the initiation of the pulse irradiation were noises caused by the pulse irradiation. 7. Post-pulse Examinations of the Test Fuel Rod After the pulse irradiation, the inner capsule which contained the test fuel rod was transported to the Research Hot Laboratory for PIEs (Post-Irradiation Examinations) consisting -16- JAERI-Research 95-087 nondestructive tests and destructive tests. The nondestructive test includes (1) visual observation and photography, (2) X-ray radiography, (3) dimensional measurement, (4) eddy current test of the test fuel rod and (5) gamma ray scanning. The destructive test includes (1) additional fission gas release during the pulse irradiation, (2) optical microscopy, (3) SEM and EPMA tests, (4) measurement of the fuel pellet density, (5) measurement of the cladding tube hardness, (6) the cladding ring tensile tests and (7) autoradiography 7.1 Non-destructive Tests (1) Visual Observation and Photography Photograph 14 shows the test fuel rod attached to the support after disassembling the capsule. The photographs were taken from the two directions of zero and 90 degrees. From this photographs, no abnormality was seen. Photograph 15 shows the appearance of the test fuel rod detached the support. Discoloration of the cladding surface due to the oxidation can be observed along the active fuel region. Though discoloration was also noted after the completion of the prc-irradiation, the change of color due to the pulse irradiation was not so significant. Slight rod bending and rod deformation, i.e. increase of the cladding diameter along the active fuel region were seen. The defects in the cladding were not observed and gap gas was kept inside the test fuel rod. (2) X-ray Radiography As shown in Photo. 16, X-ray radiographs were taken from the two directions. Any abnormalities i.e. fragment of fuel pellets were not observed in these radiographs. On the evidence of the movement marker, the axial fuel stack elongation was more than 1 mm during the pulse irradiation. (3) Dimensional Measurement The dimensional measurements of the test fuel rod were performed with the laser scanning micrometer. The examinations comprised outer diameter, bending and length measurements. The diameters were measured along two ridgc-lincs in 0=zero and 0=90 degrees over the entire axial length of the rod. The pitch of the trace was 0.5 mm, and the accuracies were ±0.005 mm -17- JAERI-Research 95-087 for the diameter and ±0.05 mm for the length. Figures 48 and 49 show the axial profile of the test fuel rod outer diameter in the fuel active region after the pulse irradiation in 6=zero and 0=90 degrees comparing with these of before the pulse irradiation, respectively. The maximum cladding diameter is 10.83 mm at 6=0 degree and z=144 mm, and 0=90 degree and z=144 mm, respectively. The maximum cladding deformation is 1.6% corresponding to that of before the pulse irradiation. The data of dimensional measurements indicated that the fuel pellets expanded or swelled to the cladding resulting in PCMI. This implies that dimensional change of the pellet arc transmitted directly to the cladding tube, and finally determine the dimensional change of the fuel rod. The total length of the test fuel rod measured in 0=0 degree and 0=90 degree directions were 299.78 mm and 299.78 mm, respectively. The rod length after the pulse irradiation is almost equivalent to that measured before the pulse irradiation. The maximum bending was approximately 0.42 mm in 0=90 degree direction and was very small. (4) Eddy Current Test As shown in Fig. 50, the eddy current test for the test fuel rod was performed with a frequency of 400 kHz and a phase angle of 30° in the horizontal and vertical directions to investigate the existence of cladding defects after the pulse irradiation. The test fuel rod had three signals on the cladding outer surface at elevations of z= around 132 mm, 150 mm and 190 mm. The signal at around 150 mm elevation was caused by the welding for the thermocouple on the cladding, and the other signals indicate existence of the minor ridge on the cladding outer surface. However, notable defects of the cladding tube were not formed during the pulse irradiation. (5) Gamma-ray Scanning Measured axial distributions of gross gamma-ray (40 to 2595 kcV) intensity for the test fuel rod is shown in Fig. 51. From the gross gamma scanning of the test fuel rod, it can be seen the gamma peaks at the pellet—pellet interfaces indicating sufficiently high pellet temperatures in that region to redistribute volatile fission products such as cesium during the pulse irradiation. As shown in Fig. 52, gamma spectrometries were conducted at the gas plenum which at z= 280 mm elevation, and the signal corresponding to krypton-85 was detected with signals of rhodium-106 and cesium-137, e.g. back ground, shown in Fig. 53. - 18 - JAERI-Research 95-087 7.2 Destructive Tests (1) Additional Fission Gas Release The test fuel rod is punctured by fission gas collection device at plenum, and free volume in the fuel rod, amounts of released gases and internal pressure are measured. Volume ratios of the released gases and abundances of isotopes are measured using the mass spectrograph. Table 15 shows test results of rod puncture and gas analysis. Fission gas inventory in the fuel rod was calculated with ORIGEN-2 code as shown in Table 16. After the pulse irradiation, krypton release of 3.8% and xenon release of 2.3% are estimated with the rod puncture, gas analysis and ORIGEN-2 code analysis. (2) Metallography Data on the fuel behavior during the pulse irradiation were obtained from metallographic and studies of fuel rod axial and vertical cross-sections. The two round slices and one vertical division, named MCI, MC2 and ML, were taken from the axial positions z = 131.5 to 141.5 mm, 181.5 to 186.5 mm and 151.5 to 171.5 mm of the test fuel rod. Sampling positions for microscopy, SEM and EPMA etc. are shown in Fig. 54. The specimens MCI and ML were subject to optical microscopy. The specimen MC2 was subject to SEM and EPMA. Photograph 17 shows macroscopic view of the specimens as polished. a) Optical Microscopy To observe fuel pellet radial and axial cross-sections and pellet/pellet boundaries, the specimens MCI and ML were subjected to optical microscopy. Enlarged photographs (400x) of the specimen ML was taken in pellet center, intermediate and outer edge regions with polished and etched surfaces, respectively, and enlarged photographs of the specimen MCI was taken in the pellet outer edge regions with polished and etched surfaces, respectively. The UO2 grain structures of the pellet specimens were revealed by dipped-ctching with a solution containing 30% H2O2 and 95% H2SO4 acid at a volume ratio of 8:1. Etching times of 90s were used. The macroscopic photograph of the specimen MCI showed that three significant crack formations occurred. It is considered that the larger cracks generated during the pre-irradiation. Other two distinct types of crack, micro cracks and short circumferential cracks, could have -19- JAERI-Research 95-087 formed during the pulse irradiation. These two cracks were chiefly confined to the fuel peripheral region where occurred large temperature gradient during the pulse irradiation. The larger cracks and the micro-cracks were also observed in the macroscopic photograph of the specimen MCI (Photo. 17). These photographs show there was any evidence of fuel melting was not observed. The enlarged photograph of the specimens MCI and ML are shown in Photos.18 and 19 with macroscopic photograph (5x), respectively. The outer edge of the fuel pellet is characterized by the loss of optically definable grain structure and an increase in porosity. Large pores were obviously present on grain boundaries and triplet points. Pore migration was not extensive and pore-free grains were also not observed. b) SEM and EPMA on the Fuel Pellet A specimen obtained from the specimen MC2, the radial cross-section at z= 181.5 mm elevation, was examined with the SEM to evaluate fuel microstructure, and the EPMA measurements were made to evaluate the distribution of retained cesium and plutonium. The SEM pictures in the center, intermediate, periphery and outer edge regions of the fuel pellet at large magnifications (lOOOx, 3000x) arc shown in Photos. 20 and 21. The SEM pictures of the pellet outer region reveals irregularly shaped intragranular porosity, a loss of the as-fabricated grain structure. Grain growth and concurrent grain boundary sweeping of fission gas bubbles have been postulated as mechanisms associated with fission gas release. For the SEM examined cross section of fuel pellet, grain boundary separation was observed in the pellet outer edge. However, the radial grain growth and intragranula bubbles were not observed. It is indicates that fission gas release during the pulse irradiation can be caused by grain boundary sweeping without grain growth. The existence of grain boundary separation and the measurement of the fractional gas release shows that gas storage within the grain is greater than migration and accumulation at the grain boundary. Distribution measured in X-ray images of cesium and plutonium is shown in Photo. 21, and anomaly and characteristic heterogeneity were not found. Generally, EPMA technic have inherent limitation in detecting quantities of fission products present. In the peripheral region of the transient tested fuel appreciable fraction of the retained Xe and Cs may be contained in gas bubbles. Since the analyzed depth in EPMA is 0.5/^m at maximum, gas in large intcrgranular bubbles is not determined. -20- JAERI-Research 95-087 (3) Measurement of the Fuel Pellet Density Density of the fuel pellet is measured by weighing a piece of pellet in the mcta-xylene under atmospheric pressure. Meta-xylene is used as substitution liquid. Bulk density of the piece of fuel pellet was 10.424 g/cm3 and 95.11%T.D. (4) Measurement of the Cladding Tube Hardness Micro-hardness measurements were taken across the cladding of the specimen MCI to study the effect of the pulse irradiation on quantitative estimation of the cladding ductility using the micro vicars hardness test device. The cladding hardness for the test fuel rod was measured on the center of the cladding wall thickness in four directions after the pulse irradiation shown graphically in Fig. 55, comparing with radial distribution of the cladding hardness for the mother fuel rod without the pulse irradiation. Hardness values were ranged from 195.8 Hv to 270 Hv after the pulse irradiation, and these were 270 Hv in averaged across the cladding wall thickness without the pulse irradiation. It is indicated that the effect of pulse irradiation on the cladding hardness was no so significantly. (5) Cladding Ring Tensile Tests The cladding ring tensile tests were performed under a cross-head speed of 0.2 mm/min at room temperature (26-28°C) in the hot cell. One round slice of the cladding tube was taken from the axial positions z = 109.5 to 124.5 mm of the test fuel rod as shown in Fig. 54. This round slice of the cladding was cut into 3 mm length for the test. Comparing to the cladding of the test fuel rod, four pieces of unirradiated cladding were made on the ring tensile tests. The specimens were tested to fracture on instron tensile machine and load-elongation data were recorded. The ultimate tensile strength and 0.2% offset yield strength were calculated from the load-elongation curves. The reductions of area and total elongation were determined from before and after test measurements. The load-elongation data of the cladding ring tensile test for the test fuel rod was shown in Fig. 56, and data of the unirradiated cladding were shown in Figs 57 through 60. Test results are summarized in Table 17. Variation of the uniform elongation was 6.0%, and it was almost the same as that of the unirradiated cladding. It is indicated that ductility of the fuel cladding in the Test MH-3 was not reduced by the pulse irradiation and also increase fuel burnup up to 39 MWd/kgU. -21- JAERI-Research 95-087 (6) Autoradiography Beta and gamma-autoradiography and a-autoradiography were performed using the specimens ML and MC2 as shown in Photos 22 and 23, respectively. Radial power skew of p y emission on the pellet was not clearly observed. Higher concentration of a emission was observed in a thin layer of the pellet edge. 8. Summary and Discussion Fuel failure was not observed in the MH tests series at the peak fuel enthalpy ranged from 47 to 67 cal/g-fuel. In the following, results obtained in the Test MH-3 will be discussed. As for fuel behavior during the pulse irradiation, the results will be compared with data obtained from the Test MH-1 and the Test MH-2. (1) Pre-irradiation Conditions The fuel rod used in the MH test series was irradiated with the fuel burnup of 38.9 MWd/kgU at the linear heat, generation rate of 19.8 kW/m during the pre-irradiation. Discoloration, absorptions of crud, scratches, speckles and grid marks on the cladding outer surface were observed. Discoloration can be thought that oxidation with particular emphasis on diffusion in both the oxides in coolant and the cladding substrate. However, significant defects of the cladding tube were not formed. The data of dimensional measurements indicated that the cladding was slightly produced creep down. The macroscopic photographs of fuel pellets showed large cracks generated during the pre- irradiation. Pores were obviously present on grain boundaries and triple points. Pore migration was not extensive and a network of intergranular porosity was not observed in all regions. For the SEM examined cross sections of fuel pellet, the radial grain growth and grain boundary separation were not observed in all regions, and development of a porous rim structure was not observed in the pellet peripheral region. Figure 61 shows a fuel burnup distribution calculated for the MH fuel by using RODBURN code(23). The RODBURN code is a combination of the fuel burnup analysis code ORIGEN and the resonance integral code RABBLE(24), and provides radially-localized burnup -22- JAERI-Research 95-087 distribution in fuels. The local bumup predicted for the MH fuel reaches 56 MWd/kgU or highcr in the outer 10 fi m region. In this region, local burnup is enhanced by plutonium production and fissioning as shown in Fig. 62. However, the structural change cannot be visually observed in the outer 10 ix m region during pre-pulse fuel examinations performed on the specimens from the MH fuel. The fractional fission gas release was 0.15% in Kr and 0.16% in Xc, respectively. According to the FASTGRASS codc(2S), the fractional fission gas release of less than 1% corresponds to less than 1000 K of fuel temperature during reactor normal operation. It shows that fission gas release during the pre-irradiation occur principally due to diffusion from the fuel pellets. (2) Transient Behavior Table 18 summarizes transient data in the Test MH-3 with the Tests MH-1 and MH-2 including the energy deposition and peak fuel enthalpy. Figure 63 shows the transient records during the pulse irradiation in the Test MH-3. Energy deposition and fuel enthalpy reached 87 cal/g "fuel and 67 cal/g "fuel, respectively. These values correspond to indices of severity of the power excursion. According to the MATPRO model(26), the peak fuel enthalpy of 67 cal/g- fuel corresponds to about 880 °C of fuel temperature. The measured cladding surface temperature was 203 ° C at maximum, and DNB occurred on the cladding surface during the transient. However, the cladding surface temperature is considerably lower than that measured in NSRR unirradiatcd fuel experiments with the same energy deposition level as shown in Fig. 64. In the Tests MH-1 and MH-2, occurrence of DNB was not observed at the peak fuel enthalpy of less than 55 cal/g "fuel. The elongation behavior of the cladding tube followed the history of the pellet stack elongation than that of the cladding temperature. It shows that the elongation of the cladding tube is chiefly caused by PCMI. The rod internal pressure immediately increased after the initiation of the pulse irradiation. This suggests that the burst release of fission gas occurred at the very early stage of the transient. Based on the local burnup information from the RODBURN code analysis, a preliminary calculation(27) regarding radial power distribution at pulse is being made with TWOTRAN code(28). Figure 65 gives a radial power distribution of the MH test fuel during pulse- irradiations. Considerably high peaking is predicted, and the radial peaking factor reaches about 2.7 at the fuel periphery. The corresponding lr.cal peak fuel enthalpy in the Test MH-3 is about 180 cal/g-fuel, and fuel pellet surface temperature is 2400 K according to the MATPRO data. -23- JAERI-Research 95-087 The fuel centerline temperature remains about 1000 K. (3) Fuel Deformation Figure 66 shows the cladding outer diameter in average of circumferential direction after the pulse irradiation in the Test MH-3 with those in the Tests MH-1 and MH-2. The cladding outer diameter of the MH-3 tested fuel rod shows the considerable increase in the whole of the active fuel regions, and it is indicated the fuel pellets expanded or swelled to the cladding resulting in PCMI. This implies that dimensional change of the pellet are transmitted directly to the cladding tube, and finally determine the dimensional change of the fuel rod and higher fission gas release. Axial deformation peaks of the test fuel rod appear at the axial center of each fuel pellet (cask-type deformation). It could be due to axial temperature distribution correlate with the pellet axial peaking. However, cladding deformations were not observed significantly after the pulse irradiations in the Tests MH-1 and MH-2. Cladding stress imposed by fuel expansion increases with peak fuel enthalpy. (4) Hydrating of the Cladding To decrease fuel cycle costs in LWRs, the continuing trend is to increase the discharge fuel burnup. Two factors limiting the discharge fuel burnup of PWR fuel are zirconium alloy cladding corrosion and hydrating. The measured hydride concentration was 41.4 ppm in entirely over the cladding wall thickness before the pulse irradiation. The migration of hydride precipitation on the outside the cladding observed from metallography, which is a sufficient condition for the enhanced corrosion, is suggested to be reduced the cladding ductility. However, the results of the cladding ring tensile tests and the cladding hardness tests did not show reducing the cladding ductility due to the increase fuel burnup up to 39 MWd/kgU and the pulse irradiation with the peak fuel enthalpy of 67 cal/g- fuel. (5) Fuel Pellet Microstructure Three types of crack formation were observed significantly on the pellet surface in MH test scries as shown in Photo. 24. Large cracks generated during the irradiation in power plants. Other two types of crack, radial micro-cracks and short circumferential cracks, were formed during the pulse irradiation. Observations of these two types of crack were principally limited to the fuel - 24 - JAERI-Research 95-087 peripheral region where large temperature gradient existed during the pulse irradiation. Generation of radial micro-cracks and short circumferential cracks was more intensive at the in- creased peak fuel enthalpy, and may contribute to additional fission gas release during the pulse irradiation. Significant fission gas release during the transient occurred due to grain boundary separa- tion, and is most dependent on the peak fuel enthalpy at pulse. For the SEM examined cross sections of the fuel pellets, grain boundary separations were observed in the pellet peripheral region in the Test MH-3, however, those were not observed in the Tests MH-1 and MH-2. The gas bubbles are individualized and lenticular at the grain boundaries and always pined by precipitates formed of metallic fission products. Because the temperature rises, the bubbles tend to coalesce and this continues until the formation of wide channels at the ternary grain boundaries. SEM pictures also show that pores were obviously present on grain boundaries and triple points, and increased in the pellet peripheral region. Pore migration was not extensive and a network of intergranular porosity was not observed in all regions in the MH test cerise. (6) Fission Gas Release The large amounts of fission products produced during the irradiation of LWR fuels have several important effects on the performance and safe operation of reactors during normal and abnormal conditions. Fission gas released from the fuel can reduce the thermal conductivity of the fuel rod fill gas and thus increase fuel temperature for a given rod power. The higher temperature increases the stored energy of the fuel rod and can cause additional release of fission products. An initial release of fission gas can also effect the subsequent release of other fission products by altering the fuel microstructure. Finally, the radioactive isotopes of the gaseous fission products make an important contribution to the radiological source term during reactor accidents. Accordingly, the important research subjects on the LWR fuel performance at extended fuel burnup is to understand the additional fission gas release behavior during the transient conditions. After the pulse irradiation, fission gas release of the test fuel rod was estimated with rod puncture, gas analysis and ORIGEN-2 code analysis. Since fission gas release during the prc- irradiation was evaluated as 0.15% in Kr and 0.16% in Xe, respectively, additional fission gas release during the pulse irradiation was 3.9% in Kr and 4.3% in Xc, respectively. Microscopic observation suggests that number of radial micro-cracks in the fuel pellet peripheral region may contribute to additional fission gas release during the pulse irradiation. It is important to know - 25 JAERI- Research 95-087 the radial distribution of retained fission products in the fuel pellet before the pulse irradiation related to the athcrmal release of fission gas. The amount of fission gas on the grain boundaries and edges depends on fuel temperature and fuel bumup(29l According to the FASTGRASS code, additional fission gas release was less than 10% shows that maximum fuel temperature was less than 2200 K, total gas release in the Test MH-3 occurred in the grain boundary and the event of the grain boundary fracturing could not be available corresponding to high fission gas release during the transient. Figure 67 shows inverse temperature dependence of krypton release for the mother fuel, the specimen MHT2, in the OGA test compared with the data for six tests in the HI serics(3U). The krypton gas release in the Test MHT2 was in reasonable agreement with the data for the HI test data and was controlled by atomic diffusion at temperatures above 1000 K. Higher krypton release (0.22% in inventory) observed at lower temperature and it was not controlled by atomic diffusion. The krypton gases released early in the OGA test period, and they can deposit or accumulate in the grain boundaries and triplet points. However, the deposited or accumulated krypton gas in the grain boundaries and triplet points was significantly lower than an amount of released krypton gas during the pulse irradiation. It is important to know gas bubble size, grain size and number density of bubbles in the vicinity of grain boundaries, and the severity of grain boundary separation during transients. 9. Conclusions One of important research programs for LWR accidents is an RIA. The third pulse irradiation test, MH-3, using Mihama unit #2 fuel with the fuel burnup of 38.9 MWd/kgU has been successfully performed to understand the fuel behavior in the NSRR. The following conclusions can be drawn from the Test MH-3. (a) The current Japanese guideline adopted a peak fuel enthalpy of 85 cal/g#fuel as a provisional failure threshold of irradiated fuel rod under RIA conditions. In test MH-3, peak fuel enthalpy reached 67 cal/g • fuel during the pulse irradiation which was less than that of the provisional failure threshold, and the test resulted in no failure with rod deformation. (b) During the pulse irradiation, the cladding surface temperature reached about 200 °C at maximum, and occurrence of DNB was observed. On the other hand, occurrence of DNB -26- JAERI-Research 95-087 was not observed in the Tests MH-1 and MH-2 with peak fuel enthalpy of less than 55 cal/g'fuel. (c) About 1.6% increase of the cladding diameter was measured through the dimensional measurement after the pulse irradiation. This deformation is relatively large comparing with those measured in the Tests MH-1 and MH-2. (d) Rod puncture and gas analysis following the pulse irradiation showed 3.8% Kr release and 2.3% Xe release. Additional fission gas release during the transient is relatively high in comparison with those during the pre-irradiation, which are evaluated as 0.15% Kr release and 0.16% Xe release. (e) Number of micro cracks and short circumferential cracks in the fuel pellet peripheral region were generated during the pulse irradiation. The crack generation could contribute to athermal release of fission gas during the pulse irradiation. Grain boundary separation was also observed in the region, and large pores were obviously present on grain boundaries and triplet points. However, gas bubble interlinkages on grain faces were not observed. (f) Variation of the cladding yield strength was less than 15%, and the cladding hardness was 1.3 times higher than that of unirradiated cladding. The hydrogen concentration was 41.4 ppm in entirely over the cladding wall thickness up to the fuel burnup of 38.9 MWd/kgU. However, these effects on the cladding tube did not induce fuel failure in the Test MH-3. Acknowledgement This work has been performed by Department of Reactor Safety Research in collaboration with Department of Hot Laboratories and Department of Chemistry and Fuel Research, JAERI. The authors would like to acknowledge and express their appreciation for the time and effort devoted by many engineers and technicians in Reactivity Accident Research Laboratory, NSRR Operation Division and Analytical Chemistry Laboratory. They are grateful to the staffs of Department of Hot Laboratories for fabricating the test fuel rod with high quality and for their careful pre- and post test fuel examinations. They also acknowledge the support and help of individuals and other organizations within JAERI too many to cite, whose contributions were critical to the success of the program. The authors greatly appreciate for Kansai Electric Power Co. and Mitsubishi Heavy Industries, Ltd. to offer the fuel rod irradiated in the Mihama unit #2 reactor. -27- JAERI-Research 95-087 References (1) Ishikawa, M. and Shiozawa, S., "A Study of Fuel Behavior Under Reactivity Initiated Accident Conditions - Review", J. Nuclear Materials, Vol.95, Nos.l&2, pp.1-30,1980. (2) e.g., Ishijima, K. and Nakamura, T., "Measurement of Transient Elongation of a Fuel Rod Under Reactivity Initiated Accident Condition and Preliminary Analysis", Proc. 1st JSME/ASME Joint Int. Conf. Nuclear Engineering (ICONE-1), Nov. 4-7, Tokyo, Japan, Vol.2, pp.263-270, 1991. (3) e.g., Fukcta, T. and Fujishiro, T., "Generation of Destructive Forces During Fuel/Coolant Interactions Under Severe Reactivity Initiated Accident Conditions", Nuclear Engineering and Design, Vol.146, pp.181-194,1994. (4) Nuclear Safety Commission of Japan, "Safety Evaluation Guidelines", 7th edition, Taisei Press, pp.239-303,1993. [in Japanese] (5) Miller, R. W.,"The Effects of Bumup on Fuel Failure: I. Power Burst Tests on Low Burnup UO2 Fuel Rods", IN-ITR-113, Idaho Nuclear Corporation, July 1970. (6) Miller, R. W., "The Effects of Burnup on Fuel Failure: Power Burst Tests on Fuel Rods with 13,000 and 32,000 MWd/MTU Burnup", ANCR-1280/TID-4500, R63, Aerojet Nuclear Company, January 1976. (7) Seiffcrt, S. L., Martinson, Z. R. and Fukuda, S. K., "Reactivity Initiated Accident Test Series: Test RIA 1-1 (Radial Average Fuel Enthalpy of 285 cal/g) Fuel Behavior Report", NUREG/CR-1465 EGG-2040, Idaho National Engineering Laboratory, September 1980. (8) Cook, B. A., Fukuda, S. K., Martinson, Z. R. and Bott-Hembree, P., "Reactivity Initiated Accident Test Series: Test RIA 1-2 Fuel Behavior Report", NUREG/CR-1842 EGG- 2073, Idaho National Engineering Laboratory, January 1981. (9) Martinson, Z. R., El-Genk, M. S., Fukuda, S. K., LaPointe, R. E. and Osetek, D. J., "Reactivity Initiated Accident Test Series: Test RIA 1-4 Quick Look Report", EGG- TFBP-5146, Idaho National Engineering Laboratory, May 1980. (10) Yanagisawa, K., Sasajima, H., Katanishi, S., Fujishiro, T., Sanpci, S., Nihei, Y., Mimura, H., Oocda, E., Yamahara, T. and Morimoto, K., "Post Irradiation Examination of 14x14 PWR Type Fuel Rod Prior to Pulse Irradiation in NSRR", JAERI-M 91-218, January 1992. (11) Fujishiro, T., Yanagisawa, K., Ishijima, K. and Shiba, K., "Transient Fuel Behavior of Per-irradiated PWR Fuels Under Reactivity Initiated Accident Conditions", J. Nuclear Materials, Vol.188, pp.162-167,1992. - 28 - JAERI-Research 95-087 (12) Nakamura, T., Yoshinaga, M., Sobajima, M., Fujishiro, T., Kobayashi, S., Yamahara, T., Sukegawa, T. and Kikuchi, T., "Experimental Data Report for Test TS-2: Reactivity Initiated Accident Test in NSRR with Per-irradiated BWR Fuel Rod", JAERI-M 93-006, February 1993. (13) Nakamura, T., Yoshinaga, M., Sobajima, M., Ishijima, K., Kobayashi, S., Yamahara, T., Sukegawa, T. and Kikuchi, T., "Experimental Data Report for Test TS-4: Reactivity Initiated Accident Test in NSRR with Per-irradiated BWR Fuel Rod", JAERI-M 94-030, March 1993. (14) Ishijima, K., Tanzawa, S., Fuketa, T., Homma, K. and Fujishiro, T., " Experimental Data Report for Test JM-1 (Series of Reactivity Initiated Accident Test in NSRR with Fuel Rod Per-irradiated in JMTR)", JAERI-M 91-127, August 1991. (15) Tanzawa, S., Fuketa, T., Homma, K., Ishijima, K. and Fujishiro, T.," Experimental Data Report for Test JM-2 (Series of Reactivity Initiated Accident Test in NSRR with Fuel Rod Prc-irradiated in JMTR)", JAERI-M 91-157, September 1991. (16) Fukcta, T., Sasajima, H., Mori, Y., Homma, K., Tanzawa, S., Ishijima, K., Fujishiro, T., Kobayashi, S., Kikuchi, T. and Sakai, H., "Behavior of Per-irradiatcd Fuel Under A Simulated RIA Condition: Results of NSRR Test JM-3 (Scries of Reactivity Initiated Accident Test in NSRR with Fuel Rod Per-irradiatcd in JMTR)", JAERI-Rcscarch 94- 006, July 1994. (17) Fuketa, T., Mori, Y., Sasajima, H. Homma, K., Tanzawa, S., Ishijima, K., Kobayashi, S., Kikuchi, T. and Sakai, H., "Behavior of Per-irradiated Fuel Under A Simulated RIA Condition: Results of NSRR Test JM-4 (Series of Reactivity Initiated Accident Test in NSRR with Fuel Rod Per-irradiated in JMTR)", JAERI-Rcsearch 95-013, March 1995. (18) Yanagisawa, K., Sasajima, H., Katanishi, S., Homma, K., Fujishiro, T., Horiki, O., Mimura H., Oeda, E., Ohwada, I., Honda, J., Yamahara, T., Ito, T., Iida, S., Takahashi, I., Sonobe, K. and Kikuchi, T., "Per-pulse Irradiation Examination, NSRR Pulse Irradiation and Post-pulse Irradiation Examination of MH-1 Fuel Rod", JAERI-M 91-220, January 1992. (19) Yanagisawa, K., Katanishi, S., Homma, K., Sasajima, H., Fujishiro, T., Horiki, O., Mimura H., Oeda, E., Ohwada, I., Honda, J., Yamahara, T., Ito, T., Iida, S., Takahashi, I., Sonobc, K., Kikuchi, T., Suzuki, T., Sonobe, T., Gunji, K., Mcguro, Y. and Adachi, T., "Pcr-pulsc Irradiation Examination, NSRR Pulse Irradiation and Post-pulse Irradiation Examination of MH-2 Fuel Rod", JAERI-M 92-015, February 1992. -29- JAERI-Research 95-087 (20) Saito, S., Inabe, T., Fujishiro, T., Ohnishi, N. and Hoshi, T., "Measurement and Evaluation on Pulsing Characteristics and Experimental Capability of NSRR". J. Nuclear Science and Technology, Vol.14, No.3, pp.226-238,1977. (21) Croff, A. G., "A User's Manual for the ORIGEN2 Computer Code", ORNL/TM-7175, Oak Ridge National Laboratory, July 1980. (22) Kanazawa, H., Sasajima, H., Homma, K., Ichise, K., Fujishiro, T. and Yamahara, T., "Fission Gas Release from High Bumup PWR Fuels Under Transient Conditions", IAEA- TECDOC-697, pp.91-98, April 1993. (23) Uchida, M. and Sato, H., "RODBURN: A Code for Calculating Power Distribution in Fuel Rods", JAERI-M 93-108, May 1993 (in Japanese). (24) Kier, P. H. and Robba, A. A., "RABBLE, A Program for Computation of Resonance Absorption in Multiregion Reactor Cells", ANL-7326, (1967). (25) J. Rest and S. A. Zawadzki, "FASTGRASS: A Mechanistic Model for the Prediction of Xc, I, Cs, Te, Ba, and Sr Release from Nuclear Fuel under Normal and Severe Accident Conditions", NUREG/CR-5840, Argonne National Laboratory Report ANL-92/3,1992. (26) Hagrman, D. L., Reymann, G. A. and Mason, R. E.(ed.), "MATPRO-VERSION 11 (Revision 2) A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior", NUREG/CR-0479, TREE-1280, Rev.2, August 1981. (27) Nakamura, T., personal communication (1995). (28) Lathrop, K. D. and Brinklcy, F. W., "Two Dimensional Multigroup Dicrete Ordinates Transport Code: TWOTRAN-II", LA-4848-MS, (1973). (29) J. Rest and G. L. Hofman, "Effect of recrystallization in high-bumup UO2 on gas release during RIA-typc transient", J. Nuclear Materials, Vol.223, pp.192-195,1995. (30) Morris F. Osborne, Jack L. Collins and Richard A. Lorenz, "Experimental Studies of Fission Product Release from Commercial Light Water Reactor Fuel Under Accident Conditions", Nuclear Technology, Vol. 78, pp.157-169, August 1987. - 30 - JAERI-Research 95-087 Table 1 The specifications of the NSRR Reactor Type Modified TRIGA-ACPR(Annular Core Pulse Reactor) Reactor Vessel 3.6m(wide)x4.5m(long)x9m(deep)open-top water pool Fuel Element 12wt% uranium-zirconium hydride(U-ZrH16) Fuel Enrichment 20wt% U-235 H/Zr atom ratio 1.6 Cladding Material Stainless steel Fuel diameter 35.6 mm Cladding diameter 37.6 mm O.D. Length of element 381 mm Number of fuel rods 157 including 8 fucl-followered control rods Control Rods Number 8 including tow safety rods Type Fuel-followcrcd Poison material Natural boron carbide(B4C) Rod drive Rack and pinion Transient Rods Number Two fast transient rods and one adjustable transient rod Type Air followered Poison material 92% enriched boron carbide(B4C) Rod drive Pneumatic and rack & pinion (Fast rods are driven only pneumatically) Experimental Cavity Inside diameter 22 cm Core Performance Steady state operation 300 kW power at maximum Pulsing operation Maximum peak power 21.1 GW Maximum core energy release 117 MJ Maximum reactivity insertion 3.4% Ak/k($4.67) Minimum period 1.17 ms Pulse width 4.4 ms at a half of the peak power Neutron life time 30//s -31- JAERI-Research 95-087 Table 2 Operation modes of the NSRR Operational Limits and Operation Modes Research items ixamples of Power History Steady State Decay Heat Operation 300kW Simulation, etc. acto r Powe en Time 1 23.000MW Natural Pulse Cold Start - up Operation 1 Energy Release 1 130MW-S RIA Reacto r Powe Time 10MW / \ Energy / \ Release Fuel Relocation / \110MW-S Coolability eacto r Po w Shaped Pulse Time Operation 10MW Energy Release /~^ 110MW-S / Power Ramping o r Powe o O / en Time 23.000MW fl Energy / Release RIA form Rated to r Powe 10MW y 1inMW., Power o o Reacto r Powe Time -32 JAERI-Research 95-087 Table 3 The specifications of the mother fuel rod Fuel Assembly Configuration 14X14 Fuel Rod Length 3856 mm Fuel Pellet Stack Length 3642 mm Fuel Weight 455 kgUO2 Fuel Rod Fuel Pellet ^U Enrichment 2.6% Diameter 9.29 mm Length 15.2 mm Density 95 % T.D. Cladding Tube Material Zircaloy-4 Outer Diameter 10.72 mm Inner Diameter 9.48 mm Wall Thickness 0.62 mm Gas Plenum Fill Gas Pressure Helium : 3.1 MPa Air : 0.1 MPa Plenum Length 179 mm -33- JAERI-Research 95-087 Table 4 Irradiation conditions of the Mihama Unit #2 Reactor Reactor Mihama Unit 2 Fuel burn-up (MWd/kgU) 38.9 1st. cycle 13.9 2nd. cycle 24.3 Linear heat rate (kW/m) 3rd. cycle 21.7 4th. cycle 19.3 Kr 0.146 Fission gas release rate (%) Xe 0.161 -34- JAERI-Research 95-087 Table 5 Results of rod puncture test for the mother fuel rod Gas volume (cm3)* 676.5 Rod free volume (cm3) 15.0 Rod internal pressure (MPa) 4.57 H2 <0.1 He 99.3 N2 <0.1 Gas content (v/o) o2 <0.1 Ar 0.44 Kr 0.053 Xe 0.56 Isotope composition (%) Kr-83 11.1 Xe-131 7.9 Kr-84 32.0 Xe-132 21.1 Kr-85 4.0 Xe-134 27.6 Kr-86 53.0 Xe-136 43.5 * : at 0°C, 1 atm. -35- JAERI-Research 95-087 Table 6 ORIGEN-2 calculation of inventories and estimated FP release for the mother fuel rod MH-3 mother fuel rod Inventory' Punctured gas Estimsted FP Release (g-atom) (g-atom) (°/o) Kr 1.093xl(T2 1.601xl0-5 0.146 Xe 1.050XKT1 1.693X10"4 0.161 Kr-83 1.220xl0"3 1.761X10-6 0.144 Kr-84 3.479xlO"3 5.122x10-* 0.147 Kr-85 6.406X10"4 6.403xl(T7 0.100 Kr-86 5.549xlO"3 8.483x10"* 0.153 Xe-1 3 1 8.176xl(T3 1.336xlO"s 0.163 Xe-1 32 2.264xlO"2 3.569xlO"5 0.158 Xe-1 34 2.9O5xlO"2 4.668xlO"5 0.161 Xe-1 3 6 4.476xlO"2 7.357xl(T5 0.164 : ORIGEN-2 code estimation -36- Table 7 Results of X-ray diffraction for the mother fuel pellet Diffraction Interplanar Miller index Lattice Lattice constant of Differences Peak No. angle spacing constant un-irradiated UO2 (20) d(pm) (h) (k) (I) a(pm) a1 (pm) a-a1 (pm) 1 28.30 315.4 1 1 1 546.3 547.6 -4 2 32.74 273.6 2 0 0 547.1 547.1 0 3 46.98 193.4 2 2 0 547.1 547.1 0 4 55.72 165.0 3 1 1 547.2 546.9 3 5 58.50 157.8 2 2 2 546.6 547.0 -4 to 6 68.66 136.7 4 0 0 546.9 547.3 -4 7 75.76 125.6 3 3 1 547.4 547.1 3 8 78.16 122.3 4 2 0 547.0 547.0 0 9 87.34 111.7 4 2 2 547.0 546.9 1 10 94.04 105.4 5 1 1 547.6 546.8 8 11 105.80 96.7 4 4 0 546.8 546.8 0 12 112.76 92.6 5 3 1 547.8 546.8 10 13 115.44 91.2 6 0 0 547.2 546.9 3 14 125.86 86.6 6 2 0 547.6 546.8 8 Table 8 The conditions of the welding for the test fuel rods Electric current of welding (A) : 55 60 • ' 1 • • • • 1 • • • • Welding time (s) : 12 50 Electric current of Back Ground (A) : 35 40 - > c - - - - - CO CO 00 l_ 13 30 Time of down slope (s) : 9 o o I 20 o Pulse width (%) : 40 10 Pulse rate (%) : 40 0 0 10 15 20 25 Fix speed (%) : 50 Time (s) JAERI-Research 95-087 Table 9 The conditions and results of filling gas and seal welding for the test fuel rods Conditions Time of evacuate the welding chamber (m) 30 Vacuum rate (ton) <0.05 Filling gas He (99.995%) Holding time in an atmosphere of helium (m) MH-1 : 120 MH-2~MH-4 : 150 Electric current of welding (A/s) 20 Results Test fuel rod No. Fill gas Pressure Inside temperature of the chamber (MPa) CO) MH-1 5.05 28.7 MH-2 5.04 27.8 MH-3 5.01 26.5 MH-4 5.02 26.8 Table 10 The conditions and results of helium leak tests for the test fuel rods Conditions Test method Vacuum-pressurization Test apparatus SHIMADZU MSE-110 Standard leak (atm4cc/s) 3.2 X10"7 Test points Top and bottom and-fittings, and tips of top end-fittings Results Test fuel rod No. Leak rate (atm#cc/s) Sensitivity (atnrcc/s) MH-1 9.2 X10"9 4.8 X10"11 MH-2 5.4 X10"8 4.5 X10"11 MH-3 1.0 XIO"7 4.3 X10"11 MH-4 7.5 X10"8 4.6 X10"11 -39- JAERI-Research 95-087 Table 11 The specifications of the test fuel rod UO2peIIet Diameter 9.29 mm Length 15.2 mm Density 95 % T.D. Enrichment 2.6 w/o Shape Dished Cladding Material Zircaloy-4 Wall thickness 0.62 mm Outer diameter 10.72 . Radial gap Pellet-cladding 0.095 mm Element Overall length 308 mm Length of UO2 pellet stack 122 mm Fill gas Helium Initial internal pressure 4.57 MPa (at 0°C) -40- Table 12 Type of artificial defects on the reference tube Artificial A B c D E F G Defects Outer Scratch Outer Scratch Inner Scratch Defect Type Ridge Ridge Hall Spot (longitude) (circumference) (circumference) > M Depth (mm) — — 0.38 — 0.18 0.18 0.22 SO Width (mm) — — — — 5.00 0.24 — to Length (mm) — — — — 0.24 5.00 5.00 en I o oo Height (mm) 0.006 0.006 — — — — — —q Hall Diameter — — 0.44 3.00 — — - (mm) Table 13 List of instrumentation Cladding surface temperature R type thermocouple Coolant water temperature K type thermocouple Pellet stack axial elongation LVDT Cladding tube axial elongation LVDT Rod internal pressure Strain gauge type sensor Capsule internal pressure Strain gauge type sensor s Capsule wall strain Strain gauge T Reactor power #1 fi fission chamber Reactor power #2 fi fission chamber CO Core energy release #1 fi fission chamber en I Core energy release #2 p. fission chamber g JAERI-Research 95-087 Table 14 Pulse irradiation condition Test No.l MH-3 MH-1 MH-2 Time and date of the pulse 2:45pm, October 31, 1990 2:35pm, November 28, 1989 2:20pm, March 8, 1990 Irradiation capsule Type X-n Capsule reactivity $-2.20 $-2.25 $-2.23 Capsule internal pressure Atmospheric Initial coolant temperature Ambient Coolant flow Stagnant Reactor operation mode NP (pulsing from zero power) Inserted reactivity $4.45 (Nominal) $3.5 (Nominal) $3.85 (Nominal) $4.44 (Bank*1) $339 (Bank*1) $3.78 (Bank'1) $4.44 (Period*2) $3.83 (Period*2) Reactor period 1.21 ms - 1.74 ms Transient rod position (TA/TB/TC) 420/DN/DN 660/DN/DN 572/DN/DN Maximum reactor power 18.0 GW 7.9 GW 11.8 GW Integrated reactor power*3 99.8 MJ 70.0 MJ 79.0 MJ Driving fuel maximum temperature 785 °C 605^ 675 "C Reactor pool water temperature- 2SS°C 23.6 "C 21.4 "C Total energy deposited to the fuel'4 87 ca!/gTucl 63 cal/g-fuel 72 cal/g-fuel Peak fuel enthalpy'' 67 cal/g-fucl 47 cal/g'fuel 55 cal/g'fucl '1 : Evaluated from corresponding regulating rod position. '2 : Evaluated from the reactor period. '3 : During initial 1 second. '4 : Radial average total energy deposition during the transient '5 : Radial average peak fuel enthalpy. -43- JAERI-Research 95-087 Table 15 Results of the test fuel rod puncture after the pulse irradiation Gas volume (cm3)* 153.96 Rod free volume (cm3) 3.79 Gas content (v/o) H2 1.52 He 94.76 N2 0.24 Kr 0.20 Xe 2.01 Others 1.28 at 0°C, 1 atm. Isotope composition (%) Kr-83 10.7 Xe-130 0.22 Kr-84 33.39 Xe-131 7.02 Kr-85 4.22 Xe-132 22.27 Kr-86 51.67 Xe-134 27.42 Xe-136 43.07 -44- JAERI-Research 95-087 Table 16 ORIGEN-2 calculation of inventories and estimated FP release for the test fuel rod MH-3 test fuel rod Inventory* Punctured gas Estimated FP Release (g-atom) (g-atom) (%) Kr 3.651X10"4 1.374xlO-5 3.763 Xe 3.529xlO"3 1.382X10"4 3.916 Kr-83 4.102xl0-5 1.471x10-* 3.586 Kr-84 1.169X10-4 4.590x10"* 3.926 Kr-85 1.942xlO"5 5.801xl0"7 2.987 Kr-86 1.865X10"4 7.103xl0"6 3.809 Xe-13 0 9.854x10-* 3.039xl0"7 3.084 Xe-1 3 1 2.748X10-4 9.698x10"* 3.529 Xe-132 7.610x10^ 3.077xl0"5 4.043 Xe-134 9.763X10"4 3.788xlO"s 3.880 Xe-13 6 1.504xl0"3 5.950xl0"5 3.956 : ORIGEN-2 code estimation -45- Table 17 Results of the cladding ring tensile tests Specimen Test Yield Load Maximum Load Rupture Load Elongation(%) Initial Tensile Rate Remarks No. Temperature / Yield Stress / Tensile Stress / Rupture Stress Cross Section (mm/min) (°C) (N/mm2) (N/mm2) (N/mm2) uniform Total (mm2) Ml R.T(27) 2793.0/76.4 2949.8/80.7 2430.4/66.5 6.0 12.3 3.73 0.2 Pulse irradiation T Ul R.T(26) 2832.2/77.3 2940.0/80.2 2283.4/62.3 6.5 12.9 3.74 0.2 I en U2 R.T(26) 2773.4/79.3 2940.0/84.0 2205.0/63.0 6.1 12.6 3.57 0.2 Unirradiated U3 R.T(28) 2783.2/76.5 2959.6/81.4 2195.2/60.4 6.2 13.3 3.71 0.5^2.0 U4 R.T(26) 2812.6/81.5 2979.2/86.4 2303.0/66.8 6.2 13.7 3.52 0.5 Table 18 Summary of the transient dada in the Test MH-3 with the Tests MH-1 and MH-2 Peak Fuel Max. Temp, of Cladding Pellet Stack Fission Gas Release Enthalpy Cladding Surface Elongation Elongation (cal/g*fuel) I \y 1 (mm) (mm) Kr Xe MH-3 67 203 1.12 1.19 3.76 3.92 o 'I MH-1 47 102 0.54 — 3.09 3.37 CO m I MH-2 55 87 0.66 0.90 — - co JAERI-Research 95-087 Capsule Hold-Down Device Control Rod Device Water Level Capsule Vertical Loading ,Strage Pit Offset Loading Neutron Detectors Neutron Radiography Room Capsule Gripping Device Fig. 1 Schematics of the NSRR - 48 - JAERI-Research 95-087 Off-set Loading Tube + + + + + OOOO + + + + + + + + + + oooooooo + + + + + + + oo®ooooo®oo + + + + + + oooooooooooo + + + }yoW9oooooo'o'o'o-°'+++ + + + +Voooq^ . f \ooocr+ + + + + O®OO + + + OO®OO(DOO®OO + + + + + + + OOOOOOOO + + + + + + + + + +OOOO+ + + + + Core Shroud + Grid Hole o Fuel Element Instrumented Fuel Element Regulating Rod with Fueled Follower Safety Rod with Fueled Follower Transient Rod A, B, C with Air Follower Fig. 2 Rod arrangement of the NSRR -49- Bottom End Fitting UO2 Pellet Cladding Spring Top End Fitting y -()- (h (h ih 00 3642 179 o e- e- 3856 t?d Unit : mm o 15.2 CD CT) Detail of Pellet Fig. 3 Design of the mother fuel rod ~ 10.90 0 = 90 degree o £ 10.72 = 0 degree > o 10.50^ 0 1000 2000 3000 4000 I en Distance from Top (mm) CO en I Fig. 4 Axial profile of the mother fuel rod diameter o OS ^ 400 Gross- d[ -|2000 g. ex o 1000 ~^ : Si t- or i - 200 0 to CD to Cs- 137 cz o o 0 O 0 1000 2000 3000 4000 Distance from Top (mm) Fig. 5 Results of gross-gamma and cesium-137 intensity for the mother fuel rod scanned by continuous measurements to I 000 CO en I o a> oo oor o 0 1000 2000 3000 4000 Distance from Top (mm) Fig. 6 Result of cobalt-60 intensity for the mother fuel rod scanned by step-wise measurement 500 (cps ) Rat e cz Count i 0 0 1000 2000 3000 4000 Distance from Top (mm) Fig. 7 Result of rhodium-106 intensity for the mother fuel rod scanned by step-wise measurement en W -£ 5000 r CD cn o I o cu CO o 1000 2000 3000 4000 Distance from Top (mm) Fig. 8 Result of cesium-134 intensity for the mother fuel rod scanned by step-wise measurement cr> i c Couniting Rate (cps; Coun' Rate Icpsl 1000 500 0 0 0 0 Fig. 10Resulto f europium-154intensityforthemothefue l rodscanne Fig. 9 .jj j.,Mi.i*.*.HiAJk''''-^«ill,i,fa2y*u*jlltJLlihlllhlJjiM..iii^ by step-wisemeasuremen t step-wise measurement Result ofcesium-137intensityforthemothefuelrodscanneb 1000 000 Distance fromTop(mm) Distance fromTop(mm) 2000 2000 3000 3000 4000 4000 T 3 > S3 I 3 ^o2 CO I o £3 31.7 594. 1156.3 1718.3 2208.1 Energy (keV) Fig. 11 Gamma-ray spectrum at the fuel active region of the mother fuel rod I CD s I 31.7 594.1 1156.3 1718.3 2208.1 Energy (keV) Fig. 12 Gamma-ray spectrum at the top plenum of the mother fuel rod Vertical Division Round Slice Round Slice for Optical Microscopy for Optical Mioroscopy, for SEM and EPMA Specimen : MML Fuel Pellet Micro-/, X-ray Diffraction, Cladding Hardness Autoradiography Specimen : MMC1 Specimen : MMC2 Measurement of dydrogen concentration on Cladding OGA Test on Fuel Pellet 3 T I 1T-1 1T-2 MHT2 MHT4 MHT3 MHT1 en I s 10 10 45 20 10 10 3.5 195.0 TOP BOTTOM 1475.0 / / / / : Observation 1280.0 unit: mm Fig. 13 Sampling positions of the destructive tests for the mother fuel rod 500 CRYSTAL : LIF CO - ex. 400 o 300 Rat e a cz 200 5 1 100 a Count i 0 1 1 500 CRYSTAL: PET en 400 oo I (cps ) 300 Rat e 200 tin g 100 Coun ' 0 90 100 120 140 160 180 200 210 Distance from Top (mm) Fig. 14 Point elemental analyses by EPMA with LiF crystal (top) and PeT crystal (bottom) on the fuel pellet of the mother fuel rod (1/2) 500 400 (cp s 300 Rat e en 200 100 Count i 0 > m 500 33 CJ1 400 to (cps ) 300 CO I o Rat e oo 200 - o 100 - O 0 60 80 100 120 160 Distance from Top (mm) Fig. 14 Point elemental analyses by EPMA with TaP crystal (top) and MyR crystal (bottom) on the fuel pellet of the mother fuel rod (2/2) JAERI-Research 95-087 200 0- 180 degree 90-270 degree in O CD 00 o O 0 (1) Gross- C\J\J 0- 180 degree 90-270 degree ex o Qi O - - cc 100 CD iti n cz =3 O O 0 (2) Cs-137 Fig. 15 Radial distribution of gross-gamma and cesium-137 intensity of the fuel pellet for the mother fuel rod - 60 - 300 1 1 i i i - > o o o IE o o f 250 —- — o X) - - o o oo Un--irradiated cladding 200 1 i i i i 0 0.1 0.2 0.3 0.4 0.5 0.6 Distance from the Cladding Inner Surface (mm) Fig. 16 Distribution of cladding hardness from the inner to outer surfaces of the mother fuel rod 5000 4000- o 3000- (f) I d 2000- w CD 3. c 1000 x 14 30 55 80 105 130 Diffraction Angle (26) Fig. 17 Results of X-ray diffraction measurements for the mother fuel pellet. 25 Center Total Counting Number : 122 Measur i ng Area : 7.40x 103/im2 20- Average Grain Size : 7.3/im Maximum Grain Size : 22.2/jm _Q £ CD •5 10 C 25 O O 0 123456789 10 1112 13 1415 16 1718 19 202122 23 242526272829 30 Grain Size (/im) Fig. 18 Grain size distribution at central region of the fuel pellet for the mother fuel rod, specimen MMC1 40 Intermediate Total Counting Number : 143 Measuring Area : 6.75x Average Grain Size : 6.3/jm 30 Maximum Grain Size : 21.6)jm IIII 0 123456789 10 1112 13 1415 16 1718 19 202122 23 24 25 26 2728 29 30 Grain Size (urn) Fig. 19 Grain size distribution at intermediate region #1 of the fuel pellet for the mother fuel rod, specimen MMC1 25 Intermediate Total Counting Number : 126 Measuring Area : 6.53x103/jm2 20 Average Grain Size : 6.9jim Maximum Grain Size : 19.8/um E C7> •j= 10 C I Z5 O o C5 —4 0 r 123456789 10 1112 13 1415 16 1718 19 202122 23 24 25 26 27282930 Grain Size Fig.. 20 Grain size distribution at intermediate region #2 of the fuel pellet for the mother fuel rod, specimen MMC1 25 Periphery Total Counting Number : 1 42 Measuring Area : 7.06x103pm2 20 Average Grain Size : 7.0pm Maximum Groin Size : 17.0/mi E Z3 CD T -E 10 o Z5 o o CO en 0 123456789 10 1112 13 1415 16 1718 19 202122 23 24 25 26 2728 2930 Grain Size (/im) Fig. 21 Grain size distribution at peripheral region of the fuel pellet for the mother fuel rod, specimen MMC1 50 Center Total Counting Number : 181 Measuring Area : 6.93x103pim2 40 Average Grain Size : 5.8/jm Maximum Groin Size : 26.0/im 30 E 13 2 T w ~ 20 to 8 o s- o CD cn I 10 o ' 1 L 123456789 10 1112 13 141516 1718 19 202122 23 24 25 26 272829 30 Grain Size (/im) Fig. 22 Grain size distribution at central region of the fuel pellet for the mother fuel rod, specimen MML 60 Intermediate Total Counting Nunber : 231 3 2 50I . , — - Measuring Area : 6.05x10 /im Average Grain Size : 5.1/jm Maxmum Grain Size : 17.4/im 40- £ 30- T I § 20 3 o CO I o 10- 0 —I 1 i i i i i i i i i i i i i i i i 1 23456789 10 1112 13 14 15 16 1718 19 202122 23 2425 26 2728 29 30 Grain Size (jum) Fig. 23 Grain size distribution at intermediate region of the fuel pellet for the mother fuel rod, specimen MML 40 Periphery Total Counting Number : 156 Measuring Area : 5.18x10"Vn2 Average Grain Size : 5.5/im 30—- Maximum Grain Size : 15.9/im — 20- > en c CO ~c ZJ o O 10- I CO 123456789 10 1112 13 1415 16 1718 19 202122 23 2425 26 2728 29 30 Grain Size Fig. 24 Grain size distribution at peripheral region of the fuel pellet for the mother fuel rod, specimen MML 1000 - o 750- CD s_ Z5 -+-J o CD 500 Q_ o E CD 250 - 0 I I I 1 I I I I 1 1 I I X I 1 I I I I I I I I I I K I I I I I I I I 0 3000 6000 9000 12000 Time (s) Fig. 25 Test conditions of the out gas analyses of the fuel pellet for the mother fuel rod, Test MHT 1 2000 - o 1500 - (D o 1000 - 0 0 5000 10000 15000 20000 25000 Time (s) Fig. 26 Test conditions of the out gas analyses of the fuel pellet for the mother fuel rod, Test MHT 2 O 1500 - —3 0 I I i I i I I I HI I I i i I I I I \\ I I I I i I I I i \\ 0 5000 WOO0 15000 20000 25000 30000 Time (s) Fig. 27 Test conditions of the out gas analyses of the fuel pellet for the mother fuel rod, Test MHT 3 - MHT4 1500 — "1400 1400 1400 1400 1400 O 1000 13 > -4—• O \ s (D [\ T Q_ E 500 \ \ CO Ul \ \ \ n i . i i |\ IIII . i i 1 1 I ... I A I 1 1 IIII 0 10000 20000 30000 Time (s) Fig. 28 Test conditions of the out gas analyses of the fuel pellet for the mother fuel rod, Test MHT 4 0.25 I I I I A O Kr 0.2 —- — Xe CD A if) O CD CD 0.15 en A ;io n Ga s Fis s 1 0.1 - D O o 1 cz A o O O "o o 0.05 - — 0 I I I I 500 600 700 800 900 # 1000 1100 Annealing Temperature (°C) Fig. 29 Results of the out gas analyses for the Test MHT 1 15 CD CD O CD 10 - c CD o 00 00 00 D LZ O 1 "o 5 - 1 c CO o en "o I o s 0 600 800 1000 1200 1400 1600- 1800 2000 2200 Annealing Temperature (°C) Fig. 30 Results of the out gas analyses for the Test MHT 2 Temperature: 800-1 400'C 2.0 I I I I A O Kr 1.5 —- A Xe — c eas e CD oISS I - 1 A if) 1.0 -- o s o T o I o o o 0.5 - D A O 0 I I I I 40 60 80 100 120 140 160 Annealing Time (m) Fig. 31 Results of the out gas analyses for the Test MHT 3 Temperature: 1400°C 3.0 CD if) O 2.0 — c CD o CD if) on > if) if) • — 1 LJL. oo - LL 1.0 ~ A CO en iona l I o oo o O t 0 0 50 100 150 200 Annealing Time (m) Fig. 32 Results of the out gas analyses for the Test MHT 4 Distance from the Top (mm) 1000 2000 3000 4000 3000 0 Q_ MH-1 MH-4 MH-2 MH-3 o) 2000 Gamma-ray P MH-3 Rod 1 2 3|4 5 6 7 8 9 10 MH-3 Test Fuel Rod ;vr: Top end fitting 2 8 9 Bottom end fitting 1708.4 mm 1830.8 mm Fig. 33 Gamma-ray profile of the mother fuel rod and cutting location of the test fuel rod for the Test MH-3 RFEF NSRR Research Hot Laboratory /'Department of \ (Department of ReactOrSafety Research) /Department of \ \Hot Laboratories/ \Hot Laboratories/ Copsuie Pre-irradiation •Loading Control Device Capsule Disassembling Pre-pulse PIEs Rod Drive r\ Mechanism Post-pulse PIEs PIEs on Fuel Semi Hot Capsule Reference Rod Transportation Cave transportation s T w s Reactor s Core i- CO I Fuel Capsule Instrumentation Assembling Fitting Device Device Instrumentation Fitting Capsule Assembling, Pulse Irradiation Fig. 34 Experimental procedure Axial elongation sensor Axial elongation sensor Pre-irradiated Rod internal for cladding for pellet stack clodding tube pressur sensor Magnetic Movement Pre-irradiated Spring adaptor iron core marker fuel pellets 3 T 122mm (Active length) Bottom end fitting eton I o 308mm (Overall length) oo Fig. 35 Schematic diagram of the test fuel rod JAERI-Research 95-087 Top z Z=91.0mm — (1708.4mm) Top of UO2 pellet stack 8 0 =0 degree between letter J and letter G Z=212.6mm— dented fuel number Bottom of UO2 pellet stack 01NJGK04 Bottom Fig. 36 Definition of coordinates for the test fuel rod -81 - MH-3 Before RIA 11.0 i i i r iiii i i i i i LULU ) 10.9 :te r cu E 10.8 o Average 15 10.72 2 10.7 Oute r T 00 CO en c CD 10.6 I add i 0 Degree o Fuel Active Region 10.5 i i t i i i 75 100 150 200 230 Distance from Top (mm) Fig. 37 Profiles of cladding outer diameter for the test fuel rod before the pulse irradiation Distance from Bottom (mm) 50 100 150 Marker (I mm/pitch) X, component Y, component X2 component T 00 CO Y2 component CD cn I Axial Position (mm) X3 component Y5 component Marker (lOOmm/pitch) Fig. 38 Result of eddy current test for the test fuel rod before the pulse irradiation Distance from Bottom (mm) 50 100 150 200 250 280 Marker (Imm/pitch) r- X, component Y, component X2 component Y2 component j component Y3 component Marker (lOOmm/pitch) Fig. 39 Result of eddy current test for the reference tube with artificial defects Cladding Elongation Sensor s Supporting Rig m Pellet Outer Capsule Elongation Senser Test Fuel Test Fuel o CM Inner Capsule oo en Thermocouple CO I o oo Capsule T/C for Water Pressure Sensor Temperature Fuel Pressure Sensor (Unit: mm) Instrumentation Capsule Fig. 40 Capsule for the pulse irradiation JAERI-Research 95-087 Cladding X elongation sensor •X X X 10mm ni X X Pellet stack X X elongation sensor X X o a -4—' T/C for water if) temperature Core center E E T/C for cladding CM surface temperature Fuel internal pressure sensor Fig. 41 Instrumentation for the pulse irradiation -86- JAERI- Research 95-087 1 1 ' t ' 1 I 18 GW *-r\ MH-3 U fission chamber #1 _ 15 \ 8 O 10 75 ms CL J I ' o a 5 . / - \ n , . y . \—^i i 1 i 0.18 0.19 0.20 0.21 0.22 Time (s) 0.18 0.19 0.20 0.21 0.22 Time (s) Fig. 42 Transient records of the NSRR reactor power -87- JAERI-Research 95-087 120 1 i 1 100- 100.4 - C7> (D C LJ MH-3 99.1 fission chamber #2 0 Fig. 43 Transient records of the core energy release measure by fission chambers - 88 - Cladding Surface Cladding Surface Temperature (°C) Temperature (°C) 0Q o o O Q 5' 8 s T 3 w OO s to s O) c CO l-t en CD I CL CO c •-i OQ 5" CD en a. 100 > El 50 ID O CD I O 23 0 Fig. 45 Coolant water temperature during the pulse irradiation JAERI-Research 95-087 0 0.4 0.6 Time (s) Fig. 46 Axial elongation of the fuel stack and cladding tube during the pulse irradiation -91- JAERI-Research 95-087 O CL (D CO to CD CL "5 c CD T3 O 0 0.2 0.4 0.6 0.8 1.0 Time (s) o • i I i i o MH-3 - - CD 6 • i ,.^..1 , , .. i.jnV'ni'iur'i1"1 iH.r. in •»•«»»•» li.l m 1111/V *"<••»» i> "i.v>vw •««••>< »nmii«^"l "I -^ »»»«M»»»MIM>I » «^^>« "•»»k< cn CD 4 CL "5 c 2 • - T3 O n • 1 1 1 1 0 4 6 8 10 Time • (s) Fig. 47 Fuel rod internal pressure during the pulse irradiation -92- = 0 degree 11.0 10.9 After Pulse S3 to o 10.6 Fuel Active Region 10.5 i \ i i I i 75 100 150 200 230 Distance from Top [mm] Fig. 48 Profiles of cladding outer diameter after and before the pulse irradiation in 6 = 0 degree 6 = 90 degree n.o | i r i —r 10.9 After Pulse (D 10.8 > 53 ID I CJl o 10.6 I C3 CO Fuel Active Region 10.5 75 100 150 200 230 Distance from Top [mm] Fig. 49 Profiles of cladding outer diameter after and before the pulse irradiation in 9 = 90 degree o" e- 8 21 18 10 20 121.6 20 50 1 2 307.6 > to I CD cn I s Unit: mm Fig. 50 Result of eddy current test for the test fuel rod after the pulse irradiation JAERI- Research 95-087 After pulse irradiation Top Bottom 1 2 3 4 5 6 7 8 x 1O4 7r to Q. Q> 4 en en a 0 70 100 150 200 250 Distance from Top (mm) Fig. 51 Results of gamma-ray scanning for the test fuel rod after the pulse irradiation 96 - JAERI-Research 95-087 PLENUM 600000 SEC. LIVE TIME 6E5745 SEC. TRUE TIME lO7r 5s ii 10= 10- 10* 10u I I I I 0.12 266.42 532.73 799.04 1065.35 ENERGY (Kev) Fig. 52 Result of gamma-ray measurement at plenum of the test fuel rod after the pulse irradiation BACK GROUND 50000 SEC. LIVE TIME 50010 SEC. TRUE TIME 106 r .10" z o en glO3 en I— 10" 101 0.12 266.42 532.73 799.04 1085.35 ENERGY (Kev) Fig. 53 Result of gamma-ray measurement in the background -97- Round Slice Vertical Division for Optical Microscopy for Optical Microscopy, Cladding Hardness Autoradiography Specimen : MC1 Specimen : ML Cladding Ring SEM and EPMA Tensile Test on Fuel Pellet, Autoradiography Density Measurement Specimen : MC2 5 on Fuel Pellet T CD SANA I \\\\\ : Observation o iiiliiiiLiUliinliiiilNiilmiliniliiiiHiiiLliiiiliHiliiiiiiiiiliiiiliiJiliiiiliHiliiiilii oo Top Bottom 91 121.6 85 297.6- unit:mm Fig. 54 Sampling positions for microscopy, SEM and EPMA etc. after the pulse irradiation JAERI- Research 95-087 300 O o O o co 250 Un-irradiated cladding CO CD with energy deposition O of 1 40 cal/g*U02 c o HI Un-irradiated cladding en O •I 200 "O o Without Pulse _o O Before Pulse o 150 0.0 0.1 0.2 0.3 0.4 . 0.5 0.6 Distance from the Cladding Inner Surface (mm) Fig. 55 Results of cladding hardness measurements for the test fuel rod after the pulse irradiation comparing with that of the mother fuel rod without the pulse irradiation -99- JAERI- Research 95-087 Specimen : M1 5000 -T T—i y -r I I 1 i I I Temperature : 27°C 4000 Pmax 2949.8N Py1 2793N 3000 Pbi 2430.4N ~§ 2000 1000 0 o / 0.5 1.0 1.5 2.0 0.6mm 1.24mm Displacement (mm) Fig. 56 Load-elongation data of the cladding ring tensile test for the test fuel rod after the pulse irradiation - loo- JAERI-Research 95-087 Specimen : U1 5000 Temperature : 26 C 4000 Pmax 2940N Py1 2832.2N 3000 i- Pb1 2283.4N "g 2000 1000 0 0 0.5 1.0 1.5 2.0 0.66mm 1.31mm Displacement (mm) Fig. 57 Load-elongation data of the cladding ring tensile test using unirradiated cladding - 101- JAERI-Research 95-087 Specimen : U2 5000 I I I I I I I Temperature : 26°C 4000 Pmax 2940N Py1 2773.4N 3000 Pb1 2205N "g 2000 1000 0 0 0.5 1.0 1.5 2.0 0.62mm 1.28mm Displacement (mm) Fig. 58 Load-elongation data of the cladding ring tensile test using unirradiated cladding -102 - JAERI-Research 95-087 Specimen : U3 5000 i i ii r Temperature : 28 C 4000 Pmax 2959.6N Py1 2783.2N 3000 Pb1 2195.2N "g 2000 1000 0 0 0.5 1.0 1.5 2.0 0.63mm 1.34mm Displacement (mm) Fig. 59 Load-elongation data of the cladding ring tensile test using unirradiated cladding - 103- JAERI-Research 95-087 Specimen : U4 5000 I I I I I • • I I I Temperature : 26°C 4000 Pmax 2979.2N Py1 2812.6N 3000 Pb1 2303N "g 2000 1000 0 0 0.5 1.0 1.5 2.0 0.6Jmm 1.38mm Displacement (mm) Fig. 60 Load-elongation data of' the cladding ring tensile test using unirradiated cladding -104- JAERI-Research 95-087 100 I 1 1 MH test fuel cr> RODBURN analysis ~o 50 Q_ Z5 c QQ 13 0 j I 0.3 0,5 1.0 Fuel Pellet Radius (r/ro Fig. 61 Radial distribution of the burnup for the JvlH fuel - 105- JAERI-Research 95-087 1E-3 o MH test fuel o RODBURN analysis o O U235 A Pu239 c o 5E-4 • Pu241 D c CD L O C O i r. ft r. r. r. r. T • O Q Q- Q B---Q-O-BQS 0 i i i i i I 0.3 0.5 1.0 Fuel Pellet Radius (r/r0) Fig. 62 Radial distributions of uranium-235, plutonium-239 and plutonium-241 for the MH fuel - 106- -ZOI - Rod Internal Pressure (MPa) Axial Elongation (mm) p b ere" Cladding Surface Temperature (°C) o o 1 1 1 1 I on - o o p oo "".•„'." CL Xn l C •^rvN. i " - <-i r - o OP > - odd i " ^ \ is e o o I a X) lO CO V' CD •;• MJ? \ IV CL o >lf. \ \ - \ iQ Q mpe r io n 9*. 3 Pr e 1 p —I o' CD A Q w Q 3 en U) 7 rti "D c* Q. CO c (D 1 S" (D O r-»- O f/) % -^ r-f r Q O JJ —% on p '>• ^r O •<• M^ f{ O bo ii (D _ ii O /^ a> vO Q //|/ io n I 1 U . i . 1 Integrated Reactor Power (MW*s) A80-96 300 I 'I r~ Thermocouple at axial center CD h 200 CD O O D p e 2 (SI *> en 100 c "O ~o Test No. 105-2 CD D CO 80 cal/g-U02 0.2 0.4 0.6 0.8 Elapsed Time (s) Fig. 64 Transient histories of cladding surface temperature measured in the Test MH-3 and un-irradiated fuel experiment #105-2-1. JAERI-Research 95-087 MH test fuel CD jn TWOTRAN analysis ZJ Q_ 2 ~o o -4—u' o o CD Q_ 0 0 0.5 1.0 Fuel Pellet Radius (r/r0) Fig. 65 Radial power profile of the MH test fuel at pulse - 109- JAERI-Research 95-087 Bottom TEST ID : MH-3 Enthalpy : 67 cal/g (10.72) O 10.7 "§ 10.6 TESTID:MH-1,MH-2 O Enthalpy: 47,55 cal/g 100 150 200 220 Distance from Top (mm) Fig. 66 Profiles of the cladding outer diameter in average of 6=0 and 0=90 degrees after the pulse irradiation in the Test MH-3 with those in the Tests MH-1 and MH-2 - no- JAERI-Research 95-087 Temperature (degree C) 2200 1800 1600 1400 1200 1000 800 1E+O E I . 1 1 I I I I IE _ A HI-6 • MH Test Data c 1E-1 - (Test MHT2) - o • HI Test Data E o HI 3 o : ~ A HI-2 HI-4. A 1E-2 k- A -t—i o HI-5 DC U 1 Illll l O 1E-3 =—- HI-1 A — o o o 1E-4 rli n 1 i - I , I 1E-5 . I.I.I 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10 10.5 10E+4/T (1/K) Fig. 67 Inverse temperature dependence of MH test data with specimen MHT2, compared with HI test- data(26) - in- JAERI-Research 95-087 Vertical Division (MML) 2mm Round Slice (MMC1) 2mm Round Slice (MMC2) 2mm Photo. 1 Samples for microscopy obtained from the mother fuel rod without pulse irradiation - 112- JAERI- Research 95-087 Photo. 2.1 Fuel pellet micro-structures observed in axial cross section of the mother fuel rod (specimen MML) without pulse irradiation (200x magnification) - 113 ~ 114 - JAERI-Research 95-087 Periphery Intermediate Center Polished 50/im Etched Photo. 2.2 Fuel pellet micro-structures observed in axial cross section of the mother fuel rod (specimen MML) without pulse irradiation (400x magnification) - 115 ~ 116 - Polished Pellet outer edge I o CD Photo. 2.3 Fuel pellet micro-structures observed in axial cross section of the mother fuel rod (specimen MML) without pulse irradiation, showing pellets and cladding interface (400x magnification) JAERI-Research 95-087 Periphery Intermediate Center «=s>V?/ . Polished mmm r 7^ Etched 30/im Photo. 2.4 Fuel pellet micro-structures observed in axial cross section of the mother fuel rod (specimen MML) without pulse irradiation (800x magnification) -178 - 119 ~ 120 - JAERI-Research 95-087 Edge Intermediate Center Polished 100/flTl Photo. 3.1 Fuel pellet micro-structures observed in radial cross section of the mother fuel rod (specimen MMCl) without pulse irradiation (200x magnification) - 121 ~ 122 - JAERI-Research 95-087 Outer edge Etched Etched 2mm 100pcm Intertnediate-l Intermediate-2 Center Etched iiiiiili 1OO/im Photo. 3.2 Fuel pellet micro-structures observed in radial cross section of the mother fuel rod (specimen MMCl) without pulse irradiation (400x magnification) 123 ~ 124 - Etched 30/xm ->( 1111^^ ~%y 1 *ji Wl Outer Inter- > edge iS mediate -2 S3 g- t * CD I o m OO **<\, \i /X 1> ^ ~i A* ^A & K >>; '".» It IM"mB £ Center to (S3 Photo. 3.3 Fuel pellet micro-structures observed in radial cross section of the mother fuel rod (specimen MMCl) without pulse irradiation (800x magnification) JAERI-Research 95-087 Location A Center Intermediate Periphery Location B Periphery Intermediate Center lip: ^ s^ Photo. 4 SEM images in fuel pellet radial cross section of the mother fuel rod (specimen MMC2) without pulse irradiation (1/2) - 127 ~ 128 - JAERI-Research 95-087 Periphery Intermediate Center Magnification Magnification Photo. 4 SEM images in fuel pellet radial cross section of the mother fuel rod (specimen MMC2) without pulse irradiation (2/2) - 129 ~ 130 - JAERI-Research 95-087 Radial crack Magnification Magnification Magnification 10/im ^ Photo.5 SEM images shows radial crack at location TAJ of fuel pellet for the mother fuel rod (specimen MMC2) without pulse irradiation - 131- JAERI-Research 95-087 Photo.6 SEM image shows unusual crack at location fBj of fuel pellet for the mother fuel rod (specimen MMC2) without pulse irradiation - 132- JAERI-Research 95-087 SEM & Pellet periphery line analysis Oxygen Neodymium ,20/inn Ruthenium Photo. 7.1 EPMA shows oxygen potential and distribution of fuel burnup in fuel pellet peripheral region of the mother fuel rod without pulse irradiation (specimen MMC2) - 133 — LNEXTPAGE(S) left BLANK. Pellet outer edge & cladding m^^^ii^g^$^0$- n o CO 00 en Photo. 7.2 EPMA shows oxygen potential and distribution of fuel burnup in fuel pellet outer edge and cladding of the mother fuel rod without pulse irradiation (specimen MMC2) • .t- A'- I I Vs.- . > 0 = 0 degree 0 = 270 degree w •;..-;•„ \ g % I : •> iCf"" >•' . V 00 -4 GO 00 0 = 90 degree 200/irn 0 = 180 degree 200/im Photo. 8.1 Cladding micro structure shows hydride in the cladding tube of the mother fuel rod (specimen MMCl) without pulse irradiation (lOOx magnification) JAERI-Research 95-087 9 = 0 degree 50/im G = 90 degree 9 = 180 degree Photo. 8.2 Cladding micro structure shows hydride in the cladding tube of the mother fuel rod (specimen MMCl) without pulse irradiation (400x magnification) - 139 ~ 140 - JAERI-Research 95-087 Outer Surface Intermediate Inner Surface Photo. 9.1 Cladding micro structure shows hydride in the cladding tube of the mother fuel rod (specimen MML) without pulse irradiation (lOOx magnification) - 141 ~ 142 - JAERI-Research 95-087 Inner Surface 2QQ/zm Inner Surface 50/im ^ -AV-V; '*•" -J*{ *•".• ":'"-."iUV'« ! Photo. 9.2 Cladding micro structure shows hydride in the cladding tube of the mother fuel rod (specimen MML) without pulse irradiation (400x magnification) - 143 ~ 144 - JAERI-Research 95-087 mmmmm. 0 = 0 degree outer surface L,T firn. 9 = 270 degree outer surface 200/ifn 6 = 90 degree outer surface 6 = 180 degree outer surface .-fOP/^ Photo. 10.1 Thick oxidation layer in the outer and inner surface of the mother fuel rod (specimen MMCl) without pulse irradiation (lOOx magnification) - 145 ~ 146 - 0 = 0 degree outer surface Inner surface 30/im 0 = 270 degree outer surface > W S3 S CO I g 00 I 0 = 90 degree outer surface 0 = 180 degree outer surface u^. Photo. 10.2 Thick oxidation layer in the outer and inner surface of the mother fuel rod (specimen MMCl) without pulse irradiation (800x magnification) JAERI-Research 95-087 Outer Surface Outer Surface Inner Surface Magnification 2Q0(im Magnification ^ 30jim ^, Photo. 11 Thick oxidation layer in the outer and inner surface of the mother fuel rod (specimen MML) without pulse irradiation - 149 ~ 150 - ^ JAERI-Research 95-087 Without Pulse Irradiation Optical Microscopy (MMC2) /3 • r Auto- Radiograph 2 mm a Auto- Radiograph Photo. 12 Fuel pellet autoradiographies taken from radial cross section of the mother fuel rod (specimen MMC2) without pulse irradiation - 151- 5! )2 '13 14 15 16 17 18 19 to CO I o oo Photo. 13 Visual appearance of the test fuel rod before the pulse irradiation S 0° ORENTATION 180 ORIENTATION Photo. 14 Visual appearance of the test fuel rod with supporting system after the pulse irradiation Top 0° ORIENTATION CO Ul Top 180° ORIENTATION Photo. 15 Visual appearance of the test fuel rod after the pulse irradiation JAERI-Research 95-087 Top 0° ORIENTATION Top 90' ORIENTATION Photo. 16 X-ray radiograph of the test fuel rod after the pulse irradiation - 155 ~ 156 - JAERI-Research 95-087 Vertical Division (ML) 2mm Round Slice (MCI) 2mm SEM/XMA Cutting Plane (MC2) 2mm Specimen 1mm Photo. 17 Samples for microscopy obtained from the test fuel rod after the pulse irradiation - 157 - | NEXT PAGE(S)left BLANK. JAERI-Research 95-087 Photo. 18 Fuel pellet micro-structures observed in radial cross section of the test fuel rod (specimen MCI) after the pulse irradiation - 159 ~ 160 - JAERI-Research 95-087 Polished Etched 2mm 2mm Outer edge Intermediate Center-l ' i 9 *• * * - * * Polished .'V ' .1 / Etched Photo. 19 Fuel pellet micro-structures observed in axial cross section of the test fuel rod (specimen ML) after the pulse irradiation (1/4) - 161 ~ 162 - JAERI-Research 95-087 Center-2 * . •« Polished " -' • • A- I i V.' Etched Photo. 19 Fuel pellet micro-structures observed in axial cross section of the test fuel rod (specimen ML) after the pulse irradiation (2/4) - 163 - NEXT PAGE(S) left BLANK.] JAERI-Research 95-087 Outer edge Intermediate Center-1 Polished 20/im •t , •V: •• V _ / Etched Photo. 19 Fuel pellet micro-structures observed in axial cross section of the test fuel rod (specimen ML) after the pulse irradiation (3/4) - 165 ~ 166 - Photo. 19 Fuel pellet micro-structures observed in axial cross section of the test fuel rod (specimen ML) after the pulse irradiation (4/4) JAERI-Research 95-087 •Pi •Center B Periphei7 BE| mi •••HHMHOulcrcdgc^H 1mm 200AUT1 Outer edge-1 Intermediate 200/im Outer edge-2 Center 200/im Photo. 20 SEM images in fuel pellet radial cross section of the test fuel rod (specimen MC2) after the pulse irradiation (1/3) - 169 ~ 170 - JAERI-Research 95-087 Outer edge-1 Outer edge-2 Photo. 20 SEM images in fuel pellet radial cross section of the test fuel rod (specimen MC2) after the pulse irradiation (2/3) - 171 ~ 172 - JAERI-Research 95-087 Outer edge-1 Outer edge-2 10/mi Periphery 10/xm Intermediate Center Photo. 20 SEM images in fuel pellet radial cross section of the test fuel rod (specimen MC2) after the pulse irradiation (3/3) - 173 ~ 174 - JAERI-Research 95-087 Pellet outer edge SEM Cesium X-ray image Plutonium X-ray image Photo. 21 EPMA shows distribution of cesium and plutonium in fuel pellet outer edge of the test fuel rod (specimen MC2) after the pulse irradiation (1/2) - 175- JAERI-Research 95-087 Plutonium X-ray image Photo. 21 EPMA shows distribution of cesium and plutonium in fuel pellet outer edge of the test fuel rod (specimen MC2) after the pulse irradiation (2/2) - 176- JAERI-Research 95-087 After Pulse Irradiation i^i^iii&&$8^vW^ Optical Microscopy 2mm 0*7 Auto- Radiograph 2mm a Auto- Radiograph 2mm Photo. 22 Fuel pellet autoradiographies taken from axial cross section of the test fuel rod (specimen ML) after the pulse irradiation - 177- JAERI-Research 95-087 After Pulse Irradiation Optical Microscopy (MCI) 2mm f £ • 7 Auto- Radiograph 2mm a Auto- Radiograph 2mm Photo. 23 Fuel pellet autoradiographies taken from radial cross section of the test fuel rod (specimen MC2) after the pulse irradiation and that of the mother fuel rod (specimen MMC2) without pulse irradiation - 178 - Test No. : MH-1 Test No. : MH-2 Test No. : MH-3 Peak fuel enthalpy : 47 cal/g'fuel Peak fuel enthalpy : 55 cal/g'fuel Peak fuel enthalpy : 67 cal/g'fuel Polished 2mm Photo. 24 Transverse sections of the MH-1, MH-2 and MH-3 tested fuel pellets Kt « ft 32 5 S I* 12 !? fett feffii ?2 ^3* R y - A- m 5}. 8$ n tnin, h, d 10" x p •v- E 10" P ft Rl * • Sf 7 A kg ffi, » f^ ' . ' • " 2 B!f Ril I s ') -J 1- 1. L 10' 7 T fli tfi r y 7 A 1- y t 10' +' G 10' M I'M £HE y K VB 1- +: h eV ft! VI fil mol 10' a k iffl-r-tinim u 2 Jt ffi * y T 7 cd 10 -N 9 1- h 10' * da f- fla Ul 7 i* T y rad leV=1.60218xl0-"J r & Pi 7.7-7 sr I u= 1.66054x10"" kg io- f •y d lO"2 •b y •f c 10"' 'J m S3 10"' •?4 9 a H (a© si 'Mi 10"' r / n Jil 32v 12 IO- f P Hi a IV "J Hz s-< fir. 32 n 10"" 7 j. Ah f 2 10"" T 1- a — a - h y N m-kg/s A A 2 H: IE /; >< IV Pa N/m - y b a. - IV i* *•'-.( :•!>. « r.l !/ J N-m - bar (it) r. Ui 7 •v W J/s a Gal i. a.I -5 a rraK'tui&j as «s. IIK — D y •li 51 lil , •U ffi 9 C A-s - Ci : * ^ IftlililiW I985«l -Plirlcj;5o fcfc'L. leV ••litt. iHH: +: IV r V W/A i- y y R iii.Xf I u OWli CODATA © 1986«Pfl£S5 fl) ill '8 lil 7 r 7 K F C/V 7 K rad ^> • i ( = 10 dyn) kgf Ibf MPa( = 10bar) kgf/cm2 a tin mmllg(Torr) lbf/in!(psi) 1 0.101972 0.224809 1 10.1972 9.86923 7.50062 x 10J 145.038 9.BUoo5 1 2.20462 0.O980665 1 0.967841 735.559 14.2233 4.-1-1822 0.453592 1 0.101325 1.03323 1 760 14.6959 M Iff 1 Pa-s(N-s/mI)=10P(.i!rX)(g/(cm-s)) 1.33322x10"' 1.35951 x 10"1 1.31579x10" 1 1.93368x10" 6.69476 x 10"' 7.03070 x 10-2 6.80460 x 10" 51.7149 1 X J( = IO'erg) kgf-m kVV- h cnl(3ilrttt) Btu ft • lbf cV 1 col = 4.18605 J IV 1 0.101972 2.77778x10"' 0.238889 9.47813x10"' 0.737562 6.24150 x 10" = 4.184 J 1 9.80665 1 2.72407x10"' 2.34270 9.29487 x 10"3 7.23301 6.12082 x 10" = 4.1855 J (15 (1: 3.6 x 10' 3.67098 x 10s 1 8.59999 x 10 s 3412.13 2.65522 x 10' 2.24694x10" f 4.18605 0.426858 1.16279x10"' 1 3.96759 x 10"' 3.08747 2.61272x10" 11:1(4! IPS ( Bq Ci Gy rad C/kg Sv rem 1 2.70270 x 10' 1 100 1 3876 1 100 3.7 x 10" 1 0.01 1 2.58 x 10"' 1 0.01 1 BEHAVIOR OF IRRADIATED PWR FUEL UNDER A SIMULATED RIA CONDITION [RESULTS OF NSRR TEST MH-3]