Gamma Emission Tomography for Determining Pin-wise Percent Fission Gas Release in Fuel Assemblies at the Halden Boiling Water Reactor

Transcription

1 Gamma Emission Tomography for Determining Pin-wise Percent Fission Gas Release in Fuel Assemblies at the Halden Boiling Water Reactor Scott Holcombe 1 *, Staffan Jacobsson Svärd 2, and Lars Hallstadius 3 1 Institute for Energy Technology OECD Halden Reactor Project, P.O.Box 173, 171 Halden, Norway Tel , Fax , scott.holcombe@hrp.no 2 Uppsala University, Division of Applied Nuclear Physics, Box 16, 7120 Uppsala, Sweden 3 Westinghouse Electric Sweden AB, Fredholmsgatan 22, Västerås, Sweden ABSTRACT: Gamma emission tomography is a non-destructive measurement technique, which can be used to determine the rod-by-rod distributions of gamma-emitting fission products within nuclear fuel assemblies without the need for disassembling the fuel. A gamma emission tomography measurement system for characterizing experimental fuel assemblies is currently being constructed at the Halden Boiling Water Reactor (HBWR). The capability of this system for determining the rod-by-rod fission product distribution has been investigated through a simulation study. Measurements were simulated for an HBWR fuel assembly and the HBWR gamma tomography device for emission of gamma rays from fission products in the gas plenum and the fuel stack regions. Tomographic algorithms were applied to the simulation results to reconstruct the rod-by-rod gamma-ray source distributions in these regions. These distributions were then used to calculate %FGR for each individual fuel rod. The results indicate that the Halden gamma tomography system is capable of accurately determining the relative rod-by-rod fission product distribution in the gas plenum and fuel stack regions of HBWR fuel. Furthermore, the system can accurately determine the fission product ratios, which, in turn, enable accurate calculation of %FGR. After this nondestructive measurement, the fuel may be reintroduced into the core. KEYWORDS: Gamma Emission Tomography, Nuclear Fuel, Nondestructive Measurement, Fission Gas Release I. INTRODUCTION Determination of the content of fission products and their spatial distribution in nuclear fuel is a widely used experimental method for investigating the in-reactor behavior and performance of the fuel. Gamma spectroscopy is a standard nondestructive method used for this purpose 1. Gamma spectroscopy may be applied to fuel assemblies or to individual fuel rods, pool-side at reactor sites or in hot laboratories; however, in order to characterize individual fuel rods, the rods must be removed from their fuel assemblies. Removal of fuel rods from their assemblies is time consuming and involves risk of damaging the fuel, thus, in practice gamma spectroscopy characterization is performed only on a limited number of individual fuel rods. Gamma tomography is a method based on gamma spectroscopy and tomographic reconstruction techniques which has recently been applied to nuclear fuel assemblies 2. Gamma tomography can characterize all fuel rods in a fuel assembly simultaneously, without the need to disassemble the fuel. Since it is not limited by the same time and handling constraints as single-rod gamma spectroscopy, gamma tomography has the potential to drastically increase the amount of data which can be collected. One important fuel characteristic which is regularly investigated is the phenomenon of Fission Gas Release (FGR), which refers to the release of gaseous fission products from the fuel pellets into the fuel rod gas volume. FGR is commonly quantified as the percentage of the total amount of fission gasses produced in the fuel pellets which have been released to the free rod volume referred to as Percent Fission Gas Release (%FGR). The mechanisms of FGR are complex and are the subject of ongoing studies 3-6. Predicting FGR behavior of fuel is important for ensuring that fuel rods are operated within their design and license limits since the presence of fission gasses in the fuel rod internal gas volume negatively affects the performance of fuel rods. Specifically, FGR contributes to increasing fuel rod internal gas pressure and reduces the thermal conductivity of the fill gas. Fuel rod design codes such as the Westinghouse STAV code 7 are used to calculate (among other things) the expected %FGR for specific fuel and core designs. The calculations are periodically benchmarked to destructive and non-destructive measurements of the fission gas content in the fuel rod internal gas volume. The primary destructive method consists of removing individual fuel rods from their assemblies, transporting them to a suitable laboratory where they are subsequently punctured in a controlled setting. The gas released upon

2 puncturing is analyzed to determine the volume of free gasses and the isotopic content of the gas is determined using mass spectrometry 8. The non-destructive method which is primarily used is based on gamma spectroscopy measurements of the gamma rays emitted in the decay of radioactive fission gas isotopes contained in the fuel rod gas plenum region. While the non-destructive measurements also require removal of the fuel rod(s) from the fuel assembly, they do not necessarily require transportation of the fuel rod(s) to a laboratory since these measurements may be performed in the spent fuel pools at reactor sites. The non-destructive method has the advantage that fuel rods may be measured many times throughout their lifetime, which aids in understanding the FGR behavior over time. In this work, gamma tomography is investigated as an enhanced measurement method for determining rod-wise %FGR, which would allow for more time-efficient measurements since no dismantling of the fuel is required. The method is an extrapolation of the gammaspectroscopy-based pool-side method. Its properties for %FGR determination have been investigated using simulated measurements on an HBWR fuel assembly for the measurement geometry of a gamma tomography device currently under construction at the HBWR. II. DETERMINING %FGR IN SINGLE FUEL RODS USING THE POOL-SIDE METHOD Since the gamma tomographic method investigated in this work is an advancement of the single-rod non-destructive gamma-ray spectroscopy FGR measurement method, a brief description of the single-rod method is presented here. The single-rod method, as the name implies, is based on high resolution gamma-spectroscopy measurements performed on individual fuel rods. These measurements may be performed in hot-laboratories or in spent fuel pools at commercial reactor sites In the single-rod method, fuel rods are removed from the fuel assemblies and two gamma-spectroscopy measurements per fuel rod are performed: 1) a measurement is performed in the gas plenum region and 2) another measurement is performed in the fuel stack region of the fuel rod. The amount of fission gas in the fuel rod gas plenum can then be compared to the total amount of fission gas produced in the fuel pellets in order to determine %FGR for each individual fuel rod. While both long-lived and short-lived fission gas isotopes may be measured for the purpose of determining %FGR 11,12, there is much more practical experience measuring the long-lived 8 Kr, which is useful for determining the lifetime integrated %FGR in individual fuel rods. 8 Kr is produced from fission in the fuel pellets and undergoes β- decay with a half-life of years, emitting a 14 kev gamma ray in % of all decays 13. The 14 kev peak in the measured gamma-ray spectra is obscured by the Compton background, a 12 kev peak from the decay of 106 Ru and the positron-annihilation peak (11 kev), which often makes it necessary to deconvolute the peaks using e.g. the LADAKH code 10. When measuring 8 Kr in fuel rods with a plenum spring, it may be necessary to allow the fuel to decay for six months or more, such that the Compton background and the intensity in the 11 kev positron annihilation peak have sufficiently diminished. There is often a low count rate in the 14 kev peak which demands a long measurement time in order to acquire adequate statistics. Additionally it is necessary to use a High Purity Germanium (HPGe) detector in order to be able to resolve the 14 kev peak from the 11 and 12 kev peaks. The 14 kev peak from 8 Kr cannot be easily measured in the fuel stack region of the fuel rod due to its relative low activity compared to the high activity of the fuel. Instead, the fission product 137 Cs is measured in this region, since it is produced linearly with burnup and is therefore proportional to the amount of 8 Kr produced in the fuel. 137 Cs has a long half-life of years and emits gamma rays at 662 kev 13 which are easily distinguishable in measurements of the fuel stack. %FGR is calculated according to Eq. (1), as the ratio of the amount of 8 Kr in the plenum to the amount of 8 Kr produced in the fuel (which is calculated based on the measurement of 137 Cs in the fuel stack) 11. If the measurement geometry is identical for the plenum and fuel stack measurements, neither geometrical effects nor the actual measured volume need to be considered when comparing the plenum and fuel stack measurements. K e, the calibration factor is close to unity and is derived through comparison of results obtained using this nondestructive method to results obtained through destructive testing (i.e. puncturing) of the nondestructively measured rods. The nondestructive method has been employed extensively on BWR fuel rods from 8x8 and 10x10 assemblies with fission gas release fractions up to nearly 30 % 14,1. R A 7 kk 100 Fn ( z) A L 7 Y Y 7 7 7t B 7t D 1 e e (662) (662) P Q K t t e B D 1 e e (14) (14) Eq. (1) Where index refers to 8 Kr, index 7 to 137 Cs R : Fraction of produced 8 Kr released into the free rod volume [%] A : Net counts in full energy peak (14 and 662 kev, respectively) : Accumulation (live) time of plenum and reference measurement, respectively [s] F n (z) : Axial burnup form factor at position of reference 137 Cs measurement l : Plenum length [mm]

3 L : Fuel column length [mm] k k : Compression factor taking the fuel rod internal volume distribution into account including the pellet-cladding gap and plenum compression spring volume Y : Fission yield [atoms per fission event] (E) : Gamma emission yield [photons of energy E per disintegration] λ : Decay constant [years -1 ] t B : Irradiation time [years] t D : Decay time [years] α μ (E) : Photon attenuation from plenum or fuel to detector for energy E (including absorption in cladding) P μ : Attenuation of 662 kev line in fuel (reference measurements) Q μ : Attenuation of 662 kev line in additional absorber (reference measurements) η(e) : Detector full peak efficiency at energy E K e : Calibration factor (called the Empirical correction factor in e.g. Ref. 12) III. GAMMA TOMOGRAPHY FOR IRRADIATED FUEL ASSEMBLY CHARACTERIZATION Gamma tomography, specifically Single Photon Emission Computed Tomography (SPECT), is a technique previously demonstrated for characterizing nuclear fuel assemblies 2. The technique involves two steps: first the gamma radiation field surrounding the fuel assembly at a selected axial position is recorded, and second, tomographic reconstruction techniques are used to obtain the gamma-ray source distribution within the fuel assembly cross section at the measured axial position. The gamma-ray spectroscopy system used to record the gamma radiation field surrounding the fuel may use one or several detectors. The detector(s) must be collimated to ensure that they record gamma rays emitted from a well-defined volume of the fuel. The detector(s) and fuel are translated and rotated relative to each other at a selected axial location and a gamma-ray spectrum is collected at each detector location. Spectroscopic analysis of selected peaks, allows for specific attributes of the fuel to be characterized. Tomographic reconstruction techniques are then applied to the recorded data to reconstruct the spatial distribution of the gamma-ray source within the fuel assembly. Several tomographic reconstruction techniques have been developed since the basic principles were first described by Radon in In medical applications, analytic techniques are primarily used, based on Radon or Fourier transform methods, but algebraic techniques have been judged to be better suited for heterogeneous objects such as nuclear fuel assemblies because of their capability to incorporate detailed modelling in the definition of a set of equations describing the detector response in all data points 2. III.1. Previous Tomographic Measurements on Nuclear Fuel Assemblies The gamma tomography technique has been previously demonstrated using the Uppsala University PLUTO device for measurement of rod-by-rod power distribution in commercial BWR fuel assemblies 2. In this case, the analysis was based on measurement of the 196 kev gamma rays emitted in the decay of 140 La (a daughter of 140 Ba, which has a half-life of 12.8 days), representative of the power in the last weeks of operation, and an agreement of 3.1 % between the measured and calculated relative power could be demonstrated. Algebraic reconstruction techniques were used in these reconstructions. In addition to being demonstrated for fuel performance measurements, gamma tomography has also been investigated as a safeguards verification method, which is intended to detect eventual missing or replaced rods in irradiated fuel assemblies 16,17. III.2. Gamma Tomography System at the HBWR A gamma tomography measurement system is currently under construction at the HBWR 18 and is scheduled for demonstration measurements during The system is intended for fuel performance characterization of experimental and Driver fuel irradiated in the HBWR. The design and construction of the HBWR gamma tomography system is a joint project between Westinghouse (Sweden), Uppsala University, and the Halden Reactor Project. The HBWR gamma tomography system utilizes a single HPGe detector to obtain the best-available energy resolution in the recorded gamma spectra. In this system, the fuel is rotated and positioned axially, while the detector and collimator are translated from side-to-side. The collimator dimensions were selected based on the measured activity in rods which have previously been individually measured using gamma spectroscopy 19. The intended measurement applications of the HBWR gamma tomography system include determination of rod-by-rod burnup and power distribution. A novel application of the technique to locate failed fuel rods in an assembly has also been simulated with positive results 20. The following sections describe a further evaluation of gamma tomography measurements of fission gasses for determination of rod-by-rod %FGR. IV. ROD-BY-ROD %FGR DETERMINATION USING GAMMA TOMOGRAPHY The principle of the proposed method for %FGR determination using gamma tomography is a combination of the single-rod method described in Section II with gamma tomography. In fact, the same comparison can be made between the 8 Kr in the gas plenum to the 137 Cs in the fuel stack; however, gamma tomography measurements are

4 instead used to obtain the rod-by-rod distribution of 8 Kr and 137 Cs. Specifically, two gamma tomography measurements should be performed for each fuel assembly: 1. Gamma tomography measurement of the gas plenum region and reconstruction of the rod-by-rod activity distribution of a selected fission gas isotope. 2. Gamma tomography measurement of the fuel stack region and reconstruction of the rod-by-rod distribution of 137 Cs. Besides 8 Kr, it has been shown that there are additional fission gas isotopes which may be measured and used to evaluate the FGR behavior of fuel rods 12. As noted in Section II there are challenges associated with measuring the 8 Kr 14 kev peak for single-rod %FGR determination, including potentially long measurement times, and the interfering peak at 11 kev. The rods simulated in this work have been previously measured and are known to have adequate intensity in the 8 Kr 14 kev peak 19. V. SIMULATIONS In accordance with the proposed method, gamma tomography measurements were simulated, using the Monte Carlo code, MCNP 21. The simulations corresponded to measurement of 8 Kr in the plenum region and measurement of 137 Cs in the fuel stack region of the fuel. Measurements were simulated over 120 evenly distributed angular positions surrounding the fuel, i.e. at an angular step-size of 3. The number of lateral measurement positions was 17 and 33, respectively, for the plenum and fuel stack measurements. V.1. Fuel Assembly Models Two fuel geometries were modeled: 1) the fuel stack region of the fuel where the fuel rods contain UO 2, and 2) the plenum region where the fuel rods contained Helium gas and plenum springs. In both cases, the fuel assembly was a nine-rod HBWR Driver fuel assembly. Figure 1 shows the cross section of the modeled fuel geometry. The fuel consists of nine fuel rods with Zircaloy-4 cladding, four solid Zircaloy-4 tie rods, and a stainless steel shroud. Air filled the remaining volumes in the simulations (i.e. between the fuel rods inside the stainless steel shroud as well as outside the fuel and in the collimator slits). For the fuel stack model the rods were filled with UO 2 and for the plenum model, the rods were filled with Helium gas and plenum springs which were approximated as a series of tilted Inconel-x70 rings inside each plenum. Figure 1, Cross-section view of a nine-rod HBWR Driver fuel assembly. V.2. Simulated Gamma-ray Source Since an HPGe detector, offering high energy resolution gamma-ray spectroscopy, will be used in the HBWR gamma tomography device, full-energy gamma-ray transport was of primary interest. Accordingly, the simulated sources were mono-energetic 14 kev gamma rays in the plenum region and mono-energetic 662 kev gamma rays in the fuel stack region. For the plenum region and fuel stack region simulations, the rod-by-rod source distribution was uniform. In order to reduce simulation time, Source Biasing was used to direct the emitted gamma rays preferentially within a narrow conical volume towards the detectors. For each detector position, the detected gamma rays were tallied according to the fuel rod from which they were emitted, whereby each fuel rod s contribution factor to the detector tally at each lateral and angular measurement position was obtained. The simulated measurement results could thus easily be adjusted to correspond to any arbitrary rod-by-rod source distribution by simply multiplying the desired rod-by-rod sources by their corresponding contribution factors. The fuel rods were positioned in their as-designed geometry and the fuel assembly was positioned with its center at the center of rotation in the measurement system. Methods of compensation for eventual displacement of the fuel assembly or individual fuel rods have been previously developed 22,23, and in order to simplify the simulations these effects were not considered in this work. V.3. Modeled Instrumentation Tie rod Fuel rods Shroud Collimator slits The modeled instrumentation was based on the HBWR gamma tomography device which is currently under construction 18. In order to reduce simulation time the collimator material surrounding the collimator slit was truncated such that all lateral measurement positions for each angular position could be simulated simultaneously. Figure 1 shows the collimator slits at all lateral positions at the 0 orientation. Two sets of simulations were performed where the collimator material was first modeled as ideal (i.e. only gamma-ray transmission through the collimator slit was assumed), and secondly, as the actual material used in the HBWR device, which in-turn yields simulation results which account for full-energy gamma-ray transmission through the

5 collimator material. The detectors were modeled as a surface tally at the rear opening of each collimator slit, and as such, the detector response was not included in the MCNP simulations. The length of each collimator was 7. cm, and the distance from the center of the fuel to the detectors was 84 cm. For the plenum simulations, 17 collimator slits were modeled. Each slit was 3 mm in width and 30 mm in height, with a lateral step size of 4 mm. For the fuel stack simulations, 33 collimator slits were modeled where each slit was 1 mm wide, 22 mm in height, and with a lateral step size of 2 mm. V.4. Simulated Projections At each angular step, the proportion of gamma rays emitted from each rod which passed through the collimator slits to reach the detectors with their full energy was tallied. The tally results from each angular step form a projection, such as the one shown in Figure 2 which corresponds to a simulated fuel measurement at the 0 orientation (i.e. the geometry shown in Figure 1). from individual rod-by-rod gamma spectroscopy measurements (described in Ref. 19). Table 1 shows the calculated emission rate per cubic centimeter for each of the nine rods in the plenum region and the fuel-stack region, as calculated from the measurement results in Ref. 19. The calculations were based on the measured count rates for the 14 kev and 662 kev peaks in each rod, and corrections were made for detection efficiency and for the gamma ray attenuation between the source and detector in the actual measurement geometry. In the simulations, a homogeneous internal source distribution in each rod was used. Table 1, Selected source distribution for the Emissions [s -1 cm 3 ] 662 kev 14 kev gamma rays in gamma rays in Rod the fuel stack the plenum Rod number (see Ref. 19) E E E E E E E E E E E E E E E E E E Counting statistics were not considered in the simulated measurements or tomographic reconstructions. Uncertainties in the simulated measurements (i.e. in the MCNP tally results) were less than 1% for the tally results which were not close to zero. VI. RESULTS VI.1. Tomographic reconstructions Figure 2, Example projection from the 0 rotation angle. As described in Section V.2., the simulated rod-to-rod source distribution was uniform; however, the tallies were performed in such a way that adjustment of the projections according to any arbitrary source distribution was possible. For ease of discussion, the fuel rods were numbered according to Figure 3. Figure 3, The fuel rod numbering scheme. The rod activities used to create the simulated projections correspond to the actual rod activities determined The software used to perform the reconstruction was the same software developed for the device described in Ref. 2 and which was used to reconstruct the rod-by-rod 140 Ba/ 140 La distribution in commercial BWR fuel assemblies. The geometry for the fuel model described in Section V.1 was given as input in the tomographic reconstruction software along with the material densities of the fuel and linear attenuation coefficients for the 14 kev and 662 kev gamma rays in the different materials. Since mono-energetic sources were considered in the simulations, it was not necessary to consider deconvoluting the 14 kev peak from the interfering peaks at 11 kev and 12 kev. In the case of actual tomographic measurements of the 14 kev peak, this must be considered and may be accomplished using e.g. the LADAKH code 10. The tomographic reconstruction software utilizes the Algebraic Reconstruction Technique (ART) to reconstruct the rod-by-rod activity distribution for a set of measurements (which correspond to a selected axial location). A thorough description of the method implemented in the software is found in Ref. 2. Tomographic reconstructions were performed for a uniform rod-to-rod activity distribution as well as the activity distribution described in Table 1. The reconstructed activities

6 for the uniform rod-to-rod source distributions corresponded with high precision to the simulated uniform source distribution in both the fuel and the plenum regions. The standard deviation (1σ) of the reconstructed activity distribution compared to the simulated activity distributions in the plenum and fuel were 0.074% and 0.00%, respectively, without simulated collimator transmission. When collimator transmission is simulated, the standard deviations become 0.21% and 0.03%, respectively for the plenum and the fuel regions. Table 2 and Table 3 show the reconstructed distributions for the case where the source distribution was non-uniform for the plenum and fuel stack regions, respectively. The relative standard deviation (1σ) of simulated versus reconstructed activities was 0.131% for the plenum region and 0.231% for the fuel region. Table 2, Simulated and reconstructed relative 14 kev emission rate distributions for the plenum region with non-uniform source distribution and ideal collimators. (normalized to the average). Rod Deviation (Reconstructed vs. Simulated) [%] Simulated distribution Reconstructed distribution % % % % % % % % % Table 3, Simulated and reconstructed relative 662 kev emission rate distributions for the fuel region with non-uniform source distribution and ideal collimators. (normalized to the average). Rod Deviation (Reconstructed vs. Simulated) [%] Simulated distribution Reconstructed distribution % % % % % % % % % VI.2. Determination of %FGR The calculation of %FGR using Eq. (1) is based on the relative count rates of 8 Kr in the gas plenum and 137 Cs in the fuel stack. In order to evaluate the suitability of gamma tomography measurements for determining %FGR, the relative emission rates of 8 Kr in the gas plenum and 137 Cs in the fuel stack are therefore of interest. The rod-wise, relative emission rates could be extracted from the reconstructions of the gas plenum and fuel stack for each fuel rod and a comparison of the reconstructed ratios to the simulated ratios is presented in Table 4. The reconstructed ratios were systematically less than the simulated ratios by an average factor of 2.7%. Table 4, Reconstructed ratios of the 14 kev emission rate in the plenum region to the 662 kev emission rate in the fuel stack for each rod. Rod Simulated ratio of 8 Kr/ 137 Cs Reconstructed ratio of 8 Kr/ 137 Cs Deviation (Reconstructed vs. Simulated) [%] E E % E E % E E % E E % 2.017E E % E E % E E % E E % E E % The reconstructed ratios in Table 4 were used to calculate the rod-by-rod %FGR according to Eq. (2). Specifically, Eq. (2) is a modified version of Eq. (1) where the terms in the equation in Eq. (1) corresponding to corrections for attenuation and geometrical effects, have been removed since these effects have been accounted for in the tomographic reconstructions. Note that the term accounting for relative detection efficiencies in Eq. (1) has been removed in this case since the simulated tomographic reconstructions do not consider this effect, whereas in actual measurements, relative detection efficiencies must be taken into account. The calibration factor was equal to unity in these calculations. The calculated rod-wise %FGR is shown in Table. R A 100 k F ( z) Y 7t B 7t D 1 e e K t B td e 1 e e k 7 7 n A7 L Y Eq. (2) Where index refers to 8 Kr, index 7 to 137 Cs R : Portion of produced 8 Kr released into the free rod volume [%] A /A 7 : Ratio of emission rate per cm 3 in the measured volume of 14 and 662 kev respectively F n (z) : Axial burnup form factor at position of reference 137 Cs measurement l : Plenum length [mm] L : Fuel column length [mm]

7 k k : Compression factor taking the fuel rod internal volume distribution into account including the pellet-cladding gap and plenum compression spring volume. Y : Fission yield [atoms per fission event] (E) : Gamma emission yield [photons of energy E per disintegration] λ : Decay constant [a -1 ] t B : Irradiation time [a] t D : Decay time [a] K e : Calibration factor Table, Rod-wise %FGR, calculated according to Eq. (2) and using the tomographically reconstructed rod-wise emission rates per cm 3 for 14 kev and 662 kev gamma rays from the plenum and fuel regions of the assembly. Rod Deviation (calculated vs. reference) %FGR from Ref. 19 %FGR calculated in this work % % % % % % % % % The rod-wise %FGR calculated using Eq. (2) and the tomographically reconstructed emission rates for 14 kev gamma rays in the plenum and 662 kev gamma rays in the fuel stack, are systematically lower than the corresponding rod-wise %FGR calculated using the single-rod gamma spectroscopy results and Eq. (1) by an average factor of 2.7%. This is a direct result of the relative lower reconstructed emission rates in the plenum region compared to the reconstructed emission rates in the fuel-stack region. This systematic difference may be attributed to e.g. gamma-ray transmission through the collimator material (which is currently excluded when modelling the HBWR tomographic setup in the reconstruction code) or to the different treatment of the plenum spring in the simulations and tomographic reconstructions. By including collimator transmission and/or addressing the treatment of the plenum spring, even better agreement can be expected. Another possible way to adjust for this type of systematic deviation is by use of the calibration factor K e in Eq. (2), and such a procedure may be employed when benchmarked measurement data are readily available. VII. CONCLUSIONS, DISCUSSION AND OUTLOOK Based on the simulated measurements and tomographic reconstructions performed in this work, it has been indicated that gamma tomography can reconstruct the relative rod-wise gamma-ray emission rates (and thus the relative rod-wise concentration of the respective emitting isotope) with high precision in both the plenum region and fuel stack region of an HBWR Driver fuel assembly. Tomographic reconstruction of the relative rod-wise emission rates of 8 Kr/ 137 Cs per cubic centimeter also resulted in values which were in good agreement with simulated ratios. Additional corrections to the systematic differences seen can be introduced for the experimental setup once benchmarked measurements are readily available. Using the tomographically reconstructed emission rate ratios, rod-wise %FGR was calculated based on the proven Westinghouse method, where the single-rod %FGR equation was modified for use with the tomographically reconstructed emission rates. The rod-wise %FGR calculated using the tomographically reconstructed emission rates and the modified %FGR equation agree with the corresponding rod-wise %FGR values calculated using the single-rod method within the typical margin of error for %FGR measurements. In conclusion, these results strongly support the applicability of tomography to determine distributions of fission products in HBWR fuel, as well as to determine %FGR in individual rods without the need to disassemble the fuel. Experimental investigation of these results is planned for mid-2014, where the simulated measurements described in this work will be experimentally performed on the actual fuel described in this work using the HBWR gamma tomography system. ACKNOWLEDGMENTS The authors would like to thank Vladimir Mozin (LLNL) and Steve Tobin (LANL) for valuable discussions regarding the MCNP simulations. REFERENCES 1. Post-irradiation Examination and In-pile Measurement Techniques for Water Reactor Fuels, IAEA-TECDOC- CD-163, IAEA, Vienna, Austria, (2009) 2. S. Jacobsson Svärd, A. Håkansson, A. Bäcklin, O. Osifo, C. Willman, P. Jansson, Non-destructive Experimental Determination of the Pin-power Distribution in Nuclear Fuel Assemblies, Nuclear Technology, vol. 11, nr. 1, pp , (200) 3. L. Noirot, MARGARET: An Advanced Mechanistic Model of Fission Gas Release Behavior in Nuclear Fuel, Proceedings of the 200 International Meeting on LWR Fuel Performance, Kyoto, Japan, October 2-6, (200)

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