Non-Destructive Gamma-Ray Spectrometry. on Spent Fuels of a Boiling Water Reactor*

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1 Journal of NUCLEAR SCIENCE and TECHNOLOGY, 12[11 pp (January 1975). Non-Destructive Gamma-Ray Spectrometry on Spent Fuels of a Boiling Water Reactor* Shojiro MATSUURA, Harumichi TSURUTA, Takenori SUZAKI, Hiroshi OKASHITA, Hirokazu UMEZAWA and Haruo NATSUME Japan Atomic Energy Research Institute** Received August 21, 1974 The spent fuels from the JPDR-I reactor were measured by means of a r-scanning facility installed in the fuel storage pool. The spatial distributions of the fission products (194Cs and '"Cs) were measured and analyzed in reference to the effects of control rod pattern. The ratios holding between the products of neutron capture and of direct fission ('39Cs/"7Cs and '6'Eu/137Cs) were also examined for its relevance to non-destructive burnup determination. The activity ratios of the fission products can be expressed by a linear function of burnup, provided that corrections are made to account for differences in irradiation history and for spatial variations in the neutron spectrum. KEYWORDS: BWR type reactors, JPDR reactor, gamma spectrometry, non-destructive testing, Li-drifted Ge detectors, burnup, fission products, cesium 137, gamma fuel scanning, spent fuels, uranium dioxide, cesium 134, spatial distribution I. INTRODUCTION Sufficiently accurate determination of the burnup of reactor fuel is one of the most important problems requiring consideration in the optimization of fuel cycles and for understanding the changes observed in reactor characteristics during operation. Precise measurements of burnup are usually performed by chemical analysis using a mass-spectrometer. Application of this method, however, is practically impossible for measuring burnup distribution over different parts of a reactor core, or for following the burnup history of a specified fuel element, because the chemical method inevitably entails destruction of the fuel elements and, moreover, requires laborious procedures to be followed. Several non-destructive methodsc"-'6j have been developed to overcome these limitations imposed by the destructive mode of measurement. Among these, 7-ray spectrometry of fission products is considered to be the most promising, particularly as a consequence of the improvements seen in the high-resolution detector and the development of sophisticated systems for data processing. The 7-rays of spent fuel can be nondestructively measured at the reactor site, provided a 7-beam collimator and measuring instruments. Some difficulty is encountered in keeping the geometrical conditions unchanged during measurement, and to normalize the 7-ray intensities to absolute values. In order to avoid this inconvenience, Rasmussen has proposed the adoption of a ratio between the activities of different fission products as a measure of irradiation'''. This still leaves some problems in applying the non-destructive technique to burnup evaluation of spent fuels : A boiling water This work was performed in connection with the IAEA Research Contract 1119/R1/R13. ** Tokai-mura, lbarak;-ken

2 Vol. 12, No. 1 (Jan. 1975) 25 reactor (BWR) is characterized by steam void formation and wide water gaps. This produces variations in the power distribution during operation, and the spatial variations thus caused in the neutron spectrum affect the ratios of the fission products. To gain a better understanding of these problems, T-spectrometric measurements were carried out on spent fuels from the Japan Power Demonstration Reactor (JPDR), which is a BWR. The results are analyzed in reference to the characteristics of the JPDR-I core. The eight fuel rods chosen for measurement were taken from a standard fuel assembly carrying the identification symbol A-20, and their locations in the assembly are as indicated in Fig. 2. Each rod was formed of two segments-upper and lowerhaving the same active fuel length. The fuel pellets are 12.5 mm in diameter and clad in Zircaloy-2 sheath 0.76 mm thick. The specifications of the rod and assembly are presented in Table 1, II. EXPERIMENTAL 1. JPDR-I Core The reactor is a natural circulation BWR of 45 MW thermal output, loaded with 2.6 To enriched UO2 fuels"). The plan of the core is shown in Fig. 1. The control rods are designated according to the coordinates relevant to their location, such as Al. " Co ntrol rod cell " is the term given to a group composed of one control rod and the four fuel assemblies surrounding it. Fig. 2 Locations of fuel rods chosen for measurement in assembly A-20 Table 1 Specifications for fuel rod and assembly Fig. 1 Plan of JPDR-I core The core was operated from October 1963 to September The operational history is given in Fig. 3. There was a

3 26 J. Nucl. Sci. Technol., Fig. 3 Operational history of JPDR-I shutdown of considerable length from June 1968 to June 1969 for inspection of the reactor pressure vessel. The average burnup of the core was estimated to be 4,400 MWd/ tm. During the irradiation, the assembly A-20 was kept at one of the highest power positions (see Fig. 1), and its average burnup was estimated to be 5,570 MWd/t from the operational data by means of the FLARE code". The reactor power was regulated by the four central control rods, while the twelve peripheral control rods were fully withdrawn. The central rods were grouped in 2 pairs (B2, C3) and (B3, C2). One of the pairs was kept at a constant height of about two thirds of full insertion, and the other was gradually withdrawn from full to two fifths insertion with increasing burnup. Their withdrawal pattern was periodically altered every 500 MWd/t to flatten the burnup distribution in the core. 2. Experimental Apparatus The measurement of 7-ray spectrum was carried out using a 7-scanning facility installed in the fuel storage pool of the JPDR. The facility consisted of a collimator set, a fuel guide and a 7-ray spectrometer. The fuel rod to be measured was held vertically on a lift and driven along a guide fixed on the pool wall. The displacement of the lift was determined within 1 mm from the readings of a selsyn indicator. The r- ray was collimated by a beam guide, constituted of a stainless steel tube, 20 mm in inner diameter and 5,750 mm long. The beam guide was fixed along a straight line by means of spacers holding it in a rigid pipe which penetrated obliquely through the ground from the floor of the building to the pool wall. The space outside the beam guide was filled with water to shield 7-rays. A planer-type Ge(Li) detector of 5.6 cms active volume was used. The detector axis was oriented at right angles to the 7-ray beam, so that the collimated 7-rays traversed the longer dimension of the active volume. Pulses from the detector were fed to a 1024 channel pulse height analyzer. Energy resolution of the counting system was 2.8 kev in FWHM for the 662 kev 7-rays from 'Cs. 3. Gamma-ray Spectrometry The total number of the measured positions in the assembly was 96 ; 24 of them were sectioned and subjected to destructive analysis. For each rod, twelve positions were selected for the measurement, as shown in Fig. 4. Ten-hour counting was applied on positions assigned to destructive analysis, and one-hour counting on the other positions. The g-rays in the range of 480,-4,390 kev were measured. The gross count rate was Fig. 4 Positions of measuring points along fuel rod

4 Vol. 12, No. 1 (Jan. 1975) ~1,000 cps, in which range the counting loss and spectrum distortion were negligible. No 7-rays other than from "Co (due to contamination of the pool water) were observed in the background spectrum. The computer code BOB-73' was used for analyzing the spectrum. The count rate of each nuclide was normalized to the value at June 1, III. RESULTS AND DISCUSSIONS A typical r-ray spectrum after 31-month cooling is shown in Fig. 5. Several nuclides, such as "Tu-"'121, '"Cs, "'Cs, 114Ce_144prp 154Eu and "Co, were identified. Figure 6 shows the vertical distributions of the fission products. As examples of the intensities measured on the nuclides, the data obtained on 'Pr, "'Cs and '"Eu from the two fuel rods 3A and 3C are plotted against that of "Ts in Fig. 7. In the case of ten-hour counting on the KC-1333, the statistical errors were 0.1, 0.5, 1 and 2% for the photopeaks of '"Cs (662 kev), '"Cs (796 kev), '"Pr (696 kev), and 1"Eu (1,275 kev), respectively. The agreement between the r-ray intensities of duplicated measurements on the segment was within a few percent for all photopeaks mentioned above. The migration of Cs was neglected in this experiment, since no anomaly in the vertical distributions of "'Cs and "Ts was observed in comparison with those of other fission products. 1. Control Rod Effect on Distribution of Fission-product Activity in Assembly The vertical distribution of the r-ray activities of the fuel rods depended on their position in the assembly A-20. In order to indicate the characteristics more clearly, the distributions of "'Cs and '"Cs normalized to those of the rod 3C are shown in Fig. 8. The results should form a set of lines parallel to the abscissa, if the horizontal distribution of burnup does not change along the vertical axis of the assembly. In actuality, the lines representing the rods 6A, 6F and 6C are evidently out of conformity. This indicates that the horizontal distribution of burnup in the assembly A-20 aried along the vertical axis. v During operation of the JPDR-I, the power in the central lower part of the core had been depressed due to neutron absorption by the control rods. Figure 9 shows the horizontal power distribution in the core obtained by gross 7-scanning on each as- Fig. 5 Typical 7-ray spectrum of irradiated fuel (31-month cooling)

5 28 J. Nucl. Sci. Technol., Fig. 6 Vertical distributions of fission products in individual fuel rods sembly during a performance test of the JPDR-f". Among the rods in the assembly A-20, those located on the side facing the core center had its output depressed by the effect of the control rod in the adjacent cell on the core inner side, this neighboring control rod effect being particularly strong on the lower parts of the fuel rod. It was thus revealed that the vertical distribution of power was affected by the control rod in adjoining cell, and this disproved the assumption widely adopted in reactor calculations, that the power distribution of each cell is independent of its neighbor, at least in the present case. 2. Correlation between Ratio of Fission Products and Burnup The activity ratios of fission products 134Cs/'"Cs and "Tu/'"Cs were studied in their relation to burnup. The nuclides 'Cs and '''Eu are produced by the neutron capture of "'Cs and ''Eu, respectively, which, in turn, are the direct products of fission, as is likewise '"Cs. In Fig. 10, the 1-ray intensity ratios in a rod (segments KA-1040 and KC-1333) are plotted against burnup on log-log scale. The burnup values* were determined from * In the present paper, the burnup values are all given in percentages of fissioned atoms in reference to the total initial amounts of all heavynuclides

6 Vol. 12, No. 1 (Jan. 1975) 29 Fig. 7 Relative activities of 144Pr, 134Cs and 154 Eu in reference to that of 137Cs Fig. 9 Horizontal power distributions at various vertical heights Fig. 8 Distributions of 1"Cs and 197Cs activities, normalized to fuel rod located at 3C Fig. 10 Measured activity ratios of i"cs/mcs and i"eu/mcs plotted against burnup

7 30 J. Nucl. Sci. Technol., destructive analysis of the fuel specimens for '"Cs and remaining U"". For each segment, a linear relation was obtained, with a common slope which was close to 1.0. The parallel lines separated, however, into two groups representing the upper and lower segments. The same behavior was found, as expected, on the other rods as well. Moreover, the ratios obtained from different fuel rods at various positions on the same plane fell along straight lines with a common slope of about 0.5 when plotted against 137Cs, as seen in Fig. 11. During the operation of the JPDR-I, it was observed from gross 1-scanning data that the position of the peak in the vertical power distribution shifted upward with burnup, as shown in Fig Conformably, accumulation of short-lived products should be larger in the upper than in the lower segment. Hence, the observed separation in the correlation curves can be attributed, at least in part, to the shift in power distribution. Fig Cs/'37Cs ratio against '"Cs on at various planes along height of fuel rods in assembly A-20 This phenomenon can be interpreted in terms of the shift of power distribution and the spatial dependence of the neutron spectrum. (1) Effect of Irradiation History When the power distribution in a core shifts during irradiation, the short-lived and long-lived nuclides come to differ in their distribution : The long-lived nuclides such as '"Cs and 151Eu retain their whole irradiation history, while the short-lived "'Cs and '"Ce reflect in their behavior mostly their more recent history. Fig. 12 Change of vertical power distribution in fuel assembly during operation The atomic density of each fission product was calculated to examine the dependence of fission-product accumulation on the irradiation history. The concentration of fission products was expressed by means of simultaneous ordinary differential equations of first order. The equations were solved numerically by the Hamming method. The neutron flux levels required for the calculation were computed by the FLARE code, using the monthly operational data. The calculated amounts of "'Cs and 144Ce are plotted against burnup in Fig. 13. Both

8 Vol. 12, No. 1 (Jan. 1975) 31 nuclides are produced directly from fission, and their loss due to neutron capture is negligible compared to 13-decay. The results obtained for the 30-yr""Cs fell along a single line, whereas those of the 284-d "'Ce separated into two lines, that for the upper segment being larger by up to 10% than for the lower at the same burnup. Fig. 13 Calculated amounts of "'Vs and 144Ce against burnup A similar distinction between long- and short-lived isotopes was seen also between the 2.05-yr134Cs and the 16-yr154Eu, which are neutron-capture products. The difference of ""Cs values between the upper and lower segments was about 5%, while the '54Eu did not separate very distinctly. The corresponding data obtained experimentally are shown in Fig. 7, where it is seen that, in contrast to the calculated lines, the ""Eu data for the upper and lower segments have separated fairly distinctly from each other. The '"Cs lines also are much more clearly separated between the two segments as compared with the calculated results. This means that factors other than the irradiation history also contribute to the distinction between the upper and lower segments in respect of fission product accumulation. (2) Effect of Spatial Variation of Neutron Spectrum A typical characteristic of the BWR is the spatial variation of neutron spectrum in the core due to steam void formation and the existence of the water gap. The amount of accumulated "'Cs depends mainly on the time integration of the neutron flux ; besides, the formation of "'Cs and 164Eu depends on the neutron capture cross section of "'Cs and 153Eu. The thermal cross section and the resonance integral are 30 and 450 b respectively for "'Cs'', and 450 and 1,500 b for 153Eu. If the neutron flux does not change during irradiation and the half-lives of the fission products are long enough compared with the irradiation time, the activity of a direct fission product NA, such as "'Cs, and that of a neutron capture product NB, such as ""Cs or '51Eu, are described by N A Oe Ef.0, (1) NBoc T r( j(1-2) 02, ( 2 ) where CP: Total neutron flux y5 multiplied by irradiation time, ectral averaged Sf:S fission cross section of the fuel rod, a(1->2): Spectral averaged neutror capture cross section of the direct fission product forming the capture product. Considering that these nuclides are formed separately with epithermal and thermal neutrons, the ratio NB/NA is expressed for a given burnup value, i. e. for constant S".1-0, by Sf, where:suffixes 2 and 3 denotes epithermal

9 32 J. Nucl. Sci. Technot., and thermal groups, respectively. The cross section 5.2(1-2) is obtained by same assembly. The slope of the lines, which was 0.5 in Fig. 11 came very close to 1.0 after applying the correction described above (Fig. 17). A similar result was obtained also for 1G4Eu/17Cs. where RI stands for the resonance integral in the epithermal region, and E, the cut-off energy. In a BWR, is about 0.1, 5 2(1-2)/a3(1 >2) is 2.6 for "'Cs and 0.6 for 16'Eu, and 02/03 ranges from about 1 to 2. Thus, the ratio NB/NA varies with 02/03 even for the same burnup. According to Eq. (3), the ratio of a given value (NA/NB)ti to the reference (NA/NB)0 is expressed by Division by f1 converts the measured activity ratio into the value under the same neutron spectrum at the reference point. Calculation by FLARE indicated that the void fraction in the JPDR-I core distributed from 0 to 45% in the vertical direction. The neutron flux ratio 02/03 at the fuel rod position 3C, computed by the MUSE code'', increased with void fraction from 1.1 to 1.7. In this calculation, the values of 5.53 kev and ev, were chosen for E,,, and E,,, respectively. These results are presented in Fig. 14. By using the correction factor f2f the measured ratios '4Cs/137Cs and 164Eu/17Cs shown in Fig. 10, were reduced to the value for the neutron spectrum at the reference point (3C, 415 mm). The results are shown in Fig. 15. The corresponding values for the upper and lower segments have come much closer to each other, though a small difference still remains. In addition, 02/953 varies with the location on the horizontal plane due to the existence of the water gap. As shown in Fig. 16, the values of 952/03 calculated by MUSE with 0% void and 1 MWd/t, ranged from 0.8 to 1.2 within the same horizontal plane in the Fig. 14 Calculated vertical distributions of void fraction and ratio of epithermal-to-thermal components ofneutron flux Fig. 15 Activity ratios '34Cs/197Cs and i"eu/'"cs reduced to corre - spond to reference position It is concluded from the foregoing observations that the activity ratios of fission

10 Vol. 12, No. 1 (Jan. 1975) 33 IV. CONCLUSION Fig. 16 Calculated distribution of neutron flux in assembly A-20 under no-void condition and at assembly averaged burnup of 1 MWd/t Fig. 17 Activity ratio of '"Cs/'"Cs corrected for horizontal variations in neutron spectrum products can tie expresseu uy a unear function of burnup, after corrections are made for differences in the irradiation history and for the variation of 0,103. Further detailed information is required on the cross sections of the fission products on the variations of the power distribution and on the neutron spectrum during operation, in order to interpret the disagreement still remaining. Non-destructive 1-ray spectrometry was carried out on spent fuels of the JPDR-I core using the T-scanning facility installed in the fuel storage pool. The results were analyzed in their relation to the reactor characteristics. The spatial distribution of the fission products in the fuel assembly was studied. The results show that the horizontal distribution of the burnup depended on the vertical position of the plane under observation, due to neutron absorption by the control rod in the adjoining cell. The relationship between the burnup and the activity ratio ('"Cs/'"Cs or 154Eu/137Cs) was examined. While the ratio of activities between products of neutron capture and direct fission is currently considered a good indicator of burnup, it was found from the present experiment that the ratios corresponding to a given burnup varied with the position in the assembly. In the case of "'Cs/'"Cs, for example, a difference amounting to 20% was observed between the ratios measured on the upper and lower segments of a fuel rod at the same burnup. This variation can be attributed to differences in the irradiation history and to spatial variations affecting the neutron spectrum during operation. It should hence be borne in mind that, in so far as BWR-type reactors are concerned, differences in the reactor characteristics must be taken into account in estimating burnup based on correlations between the amounts present of different fission products. Conversely, this finding has revealed the possibility of obtaining information on reactor characteristics, such as neutron spectrum variation and void distribution, from measurements of these correlations by non-destructive 7.-ray spectrometry. ACKNOWLEDGMENTS The authors are grateful to Mr. T. Suzuki and Mr. S. Okazaki for their assistance in the destructive measurement of burnup, and - 3 -

11 34 J. Nucl. Sci. Technol., to Mr. H. Ezure for his advice in the utilization of the JPDR-I operational data. Thanks are also due to Mr. M. Ishizuka for his valuable suggestions on this work. -REFERENCES (1) RASMUSSEN, N.C. : Proc. Symp. Safeguards Res. & Development, WASH-1076, p. 130 (1967). (2) HICK, H., LAMMER, M.: Proc. Symp. Safeguards Tech., Vol. I, p. 533 (1970), IAEA, Vienna. (3) DRAGNEV, T., BEETS, C.: EUR-4576e, (1971). (4) "Reactor Burn-up Physics", Proc. Panel Reactor Burn-up Physics, (1971), IAEA, Vienna. (5) BEETS, C., et al. : Trans. Amer. Nucl. Soc., 15, 673 (1972). (6) KOCH, L., CRICCHIO, A., GERIN, F.: Proc. Symp. Analytical Methods Nucl. Fuel Cycle, p. 523 (1972), IAEA, Vienna. (7) SHIMOOKE, T. : Nucl. Technol., 10, 257 (1971). (8) EZURE, H.: J. At. Energy Soc. Japan, (in Japanese), 13(12), 704 (1971). (9) Div. of JPDR, JAERI : Unpublished results. (10) DELP, D. L., at al.: GEAP-4598, (1964). (11) BABA, H., et al. : JAER I-1227, (1973). (12) NATSUME, H., at al.: Unpublished results. (13) WALKER, W. H. : A ECL-3037, (1972). (14) EZURE, H.: Private communication

Non-Destructive Gamma-Ray Spectrometry on Spent Fuels of a Boiling Water Reactor

Journal of Nuclear Science and Technology SSN: 0022-3131 (Print) 1881-1248 (Online) Journal homepage: http://www.tandfonline.com/loi/tnst20 Non-Destructive Gamma-Ray Spectrometry on Spent Fuels of a Boiling