Auto Sizing of CANDU Nuclear Reactor Fuel Channel Flaws from UT Scans.

The inspection of nuclear power plants is an essential process that occurs during plant outages. During this process, various systems are inspected, including the reactor's fuel channels to ensure that they are safe and reliable for the plant's operation. The inspection of Canada Deuterium Uranium (CANDU®) reactor pressure tubes, which are the core component of the fuel channels and house the reactor fuel bundles, is performed using Ultrasonic Testing (UT). Based on the current process that is followed by Canadian nuclear operators, the UT scans are manually examined by analysts to locate, measure, and characterize pressure tube flaws. This paper proposes solutions for the auto-detection and sizing of pressure tube flaws using two deterministic algorithms, the first uses segmented linear regression, while the second uses the average time of flight (ToF) within ±σ of µ. When compared against a manual analysis stream, the linear regression algorithm and the average ToF achieved an average depth difference of 0.0180 mm and 0.0206 mm, respectively. These results are very close to the depth difference of 0.0156 mm when comparing two manual streams. Therefore, the proposed algorithms can be adopted in production, which can lead to significant cost savings in terms of time and labor.

Characterization and mapping of the neutron fields around Bruce Power's 177Lu isotope production system.

Bruce Power operates a first-of-its-kind isotope production system (IPS) that enables continuous production of 177Lu within Canada Deuterium Uranium (CANDU) commercial power reactors. Located on the reactivity mechanisms deck of Unit 7, just outside of reactor containment but in close proximity to the primary heat transport (PHT) pumps, this facility offers unique advantages for 177Lu production. However, employees working in this area encounter a radiation hazard which consists primarily of photoneutrons. These originate from the base of the PHT pumps and are only present when the reactor is operating. This study evaluates neutron exposure at Bruce Power's IPS by using a nested neutron spectrometer (NNS) to determine the neutron energy spectra and absolute dosimetric quantities such as the ambient dose equivalent, H*(10). The results from the NNS are then compared to surveys performed by a portable neutron rem meter (Model NP-2 by Nuclear Research Corporation), routinely used by Bruce Power staff for workplace monitoring. While the Model NP-2 generally showed consistent results across locations, a 50% dose correction factor was identified when operators were harvesting 177Lu from the IPS. This finding highlights an opportunity to reduce the neutron dose that is assigned to operators when producing 177Lu.

Eye lens dosimetry in Canadian CANDU nuclear power plants based on operational dosimetric quantities Hp(10) and Hp(0.07).

In this article, we present a methodology for performing eye lens dosimetry in CANDU nuclear power plants using an existing and highly accurate Harshaw 4-element TLD-700 dosemeter. This dosemeter, which has been specially designed for Ontario Power Generation (OPG) and Bruce Power (BP), measures the deep and shallow personal dose equivalent quantities Hp(10) and Hp(0.07), respectively. Using these measured personal dose equivalent quantities and applying a beta-ray strength scaling factor to the Hp(0.07) measurement in particular, we have developed an algorithm that can be used to calculate the dose to the lens of the eye in mixed beta-gamma fields. This scaling factor has been developed and is primarily based on results obtained from extensive collaborative study, performed by Ontario Power Generation (OPG), Bruce Power (BP) and McMaster University, through Candu Owners Group (COG) support (Bohra et al., 2021; Laranjeiro et al., 2020). Furthermore, scaling factor F, also includes effects of protective glass eyewear and results from Whole body dosimetry intercomparison exercises. The algorithm to calculate eye lens dose at CANDU power plants has been developed, based on this scaling factor and operational dosimetric quantities Hp(10) and Hp(0.07).

Quantification of pure beta spectra in mixed beta gamma fields as part of eye lens dosimetry at CANDU power plants.

To address the issue of eye lens dosimetry in nuclear industry, we initiated the project to quantify the beta and gamma-ray source term in CANDU power plants and to convert this source term into dosimetric quantities of interest, such as eye lens dose and personal dose equivalents Hp(10), Hp(0.07). This way, the eye lens dose can be compared with dosimetric operational quantities to evaluate whether independent dosimetry is required for eye lens protection, or present dosimetry is adequate.