To validate the computational model of the reactor after it is shutdown, a simple experiment with irradiated fuel was conducted.
In September 2018, the reactor was not in operation for 20 days. In that time, the activity of the fuel elements inside the core became sufficiently low that the fuel could be placed in a dedicated transport cask. Removing the fuel provided an opportunity to test the assumptions concerning the delayed gamma source term and to validate the MCNP model for delayed gamma ray simulation.
Fuel element with ID number 7231 was the last one to be inserted into the core and consequently had the lowest activity. This fuel element is highlighted in JSI TRIGA core schematic depicted in Figure66.
Figure 66: Photo of the transport cask containing fuel element 7231.
The fuel transport cask, in which the fuel element was inserted, consists of 1.3 tons of lead and can contain one fuel element. It is made from lead and is encased by a layer of stainless steel. At the bottom
and the top there are two plugs preventing the fuel element from dropping out of the transport cask. The photograph of the cask is presented on Figure 4 and its schematic is shown on Figure 67.
Figure 67: A schematic side view of the fuel transport cask [36].
The cask with the inserted fuel element was placed on the floor of the reactor hall. The gamma dose rates were then measured around the cask at 23 predetermined locations. When fuel elements are inserted into the core, their angular orientation is not marked. Therefore, every time a fuel element is relocated, its angular orientation changes. This and inhomogeneous in-core flux distribution are the reason that a fuel element has a higher burnup on one side than the other [71]. Since the orientation is not followed, it is not known which side has the higher burnup. Therefore, the gamma dose rate was measured at four angular locations (every 90 degrees) to account for inhomogeneous angular burnup. Measurements were taken at different heights and distances from the cask. They were also taken just above the cask where the highest dose rate was expected. Finally, the transport cask was lifted off the ground and one measurement was taken just below. Measurement locations are shown in Figure 68.
Before any measurements were taken, the background dose rate was measured, and later subtracted from the measured results. Each measurement lasted 5 seconds. The probe used was the Automess 6150AD.
All measurements were taken in accordance to the internal procedures by the JSI accredited laboratory [60]. The results are given in Table 22. The origin of the x and y-axis is the middle of the fuel the z-axis is the reactor hall floor. The fuel mid-plane is located at the height of 45 cm.
Figure 68: Locations of dose rate measurements (only first quadrant is shown). Measurements were repeated every 90 degrees on all four sides of the cask to compensate for angular burnup. Location H is located 3 cm underneath the transport cask.
The same geometry was later modelled in a MCNP model containing only the fuel element, transport cask and reactor hall floor (Figure 69). At locations where dose rate was measured, spheres were modelled, as presented on Figure68. Virtual detectors were 10 cm in diameter with the exception of location A and H, where 4 cm diameter spheres were used. Inside each sphere gamma flux was tallied and converted into dose rate using conversion factors depicted on Figure23. The material composition was the same as that described in Appendix 1.
Figure 69: Vertical cross-section of the MCNP model containing the fuel element and transport cask.
Table 22: Measured and calculated gamma dose rates measured in the vicinity of the fuel element transport cask.
When modelling the delayed gamma spatial distribution source, the angular distribution was also taken into account. Extended MCNP model of TRIGA reactor was taken and eigenvalue calculation was performed. The fission rate distribution was calculated as a hundred vertical disks across the desired fuel element. Each disk was divided into eight angular slices. The calculated fission rate distribution for fuel element 7231 was applied to the delayed gamma source. Spectra were calculated for 20 hours after shutdown (Figure45). The most challenging aspect of the experiment was determining the activity of the fuel element. Recently activities are described in the operational history of the JSI TRIGA Mark II research reactor [71]. The operational logbooks were used as a source of information for reactor operation and fuel shuffling, needed for detailed burnup calculations. In total, 239 core configurations, from 1966 to 2018 were studied to obtain a final fuel isotopic composition and activity of fuel element 7231. The calculations were performed using the Serpent Monte Carlo neutron transport code [70], which is already established as a viable burnup code, due to its built-in burnup routine. The ENDF/B VII.1 nuclear library [67] was used. Three hundred different isotopes (fission products) were taken into account and transported from the previous fuel cycle (core configuration) into the next one. In Serpent, each fuel step was simulated by a single full power (250 kW) operation, followed by a cool-down period representing the time the reactor was in shutdown. The length of the operating step was determined through dividing released energy by the full power. It was assumed that this is good approximation since the reactor operates mostly at 250 kW. Furthermore, it was analysed by Pungerčič et al. that changes in uranium and plutonium concentration is minimal for different operating cycles (different operating times) where the same amount of energy was released [97].
With this, it was possible to calculate the activity of the specific fuel element at a specific time in the history of the reactor operation. The calculated activity for selected case (fuel element 7231 after 20
days cool-down time in September 2018) is 2.18 1011 Bq. Validating burnup calculation is challenging without the appropriate equipment (spent fuel gamma spectrometry), and other experiments must be considered. Work and the experiment presented in this thesis represents part of the verification of the burnup history calculations [90].
The MCNP calculated dose rates are presented in Table22. They indicate that the calculated values are underestimated. Since certain locations are symmetric over angle (e.g. 1, 4, 7 and 10 represent location A), the dose rate at these locations was averaged (Table23). In the last column, a comparison is made between the calculated and measured results. Figure70 shows the absolute results and later normalized results per location A for locations A – E and normalized results per location F for locations F – H.
Figure 70: Top: absolute values of calculated and measured dose rates separately for locations A – E and F – G. Bottom:
normalized results per location A and F for locations A – E and F – G, respectively.
Table 23: Comparison of calculated and measured results averaged over angle.
Locations Z
The calculated results are within the uncertainties of the measurements at locations closest to the transport cask (location A and D). The reason for higher measured dose rates at other locations could be to the backscattering of gamma rays from the reactor hall walls and the reactor body, which was not modelled and taken into account in the MCNP model of the transport cask. A large discrepancy between calculations and measurements is observed at locations F, G and H where the calculated values are six times larger than the measured ones. This is due to the assumption that delayed gamma rays are emitted only from the fuel region and not from the graphite pellet, which can be highly activated, and is located above and below the fuel pellet in the fuel element. In the model, this graphite pellet is shielding the gamma rays that are emitted from the fuel in the vertical direction. In reality, the pellet itself emits gamma rays in a vertical direction traveling out of the transport cask. Additionally, the detector gives a linear response for gamma rays with energies up to 700 keV whereas gamma rays emitted from the fuel can reach energies that are ten times higher, as depicted in Figure 45. Detector energy response function is presented in Figure 20.
It can be observed that the normalized calculated gamma dose rates follow the same distribution as the measured ones. The results are presented in the bottom part of Figure 70. In the future, the accuracy of the calculations will be improved by taking the activity of the graphite pellet and stainless steel into account. It is estimated that current results are accurate down to one order of magnitude. It is still not accurate enough to perform in-core calculations. However, the accuracy would be sufficient for performing dose rate analysis during various accident scenarios. Such analyses are concerned with dose rate estimates upon those that can be used to plan corrective actions and for the installation of additional components/systems if needed.
Another advantage of calculating dose rates over measurements is that simulations enable the determination of dose rates over a larger area. An example of this is when an irradiated fuel element is inside a transport cask, as presented in Figure 71. Using the model, larger area can be analysed to establish the minimum and maximum dose rate level quickly. In this way, it can be quickly observed which areas need improved shielding.
Figure 71: Horizontal (left) and vertical (right) cross-section of the gamma dose rate around the transport cask containing an irradiated TRIGA fuel element. Results are normalized per source particle (Sv/h·Bq). Fuel element activity is 2.18·1011 Bq.