Spent fuel pool

Summary

Spent fuel pools (SFP) are storage pools (or "ponds" in the United Kingdom) for spent fuel from nuclear reactors. They are typically 40 or more feet (12 m) deep, with the bottom 14 feet (4.3 m) equipped with storage racks designed to hold fuel assemblies removed from reactors. A reactor's local pool is specially designed for the reactor in which the fuel was used and is situated at the reactor site. Such pools are used for short-term cooling of the fuel rods. This allows short-lived isotopes to decay and thus reduces the ionizing radiation and decay heat emanating from the rods. The water cools the fuel and provides radiological protection from its radiation. Pools also exist on sites remote from reactors, for longer term storage such as the Independent Spent Fuel Storage Installation (ISFSI), located at the Morris Operation, or as a production buffer for 10 to 20 years before being sent for reprocessing or dry cask storage. While only about 20 feet (about 6 m) of water is

Official source

https://en.wikipedia.org/wiki/Spent_fuel_pool

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High-Fidelity Determination of Nuclide Inventories for Radioactive Waste Disposal

Valentyn Bykov

The general license application (RBG) for the Swiss deep geological repository will be submitted in 2024. Furthermore, the decommissioning of the nuclear power plants (NPPs) is coming soon, with the Mühleberg NPP (KKM) scheduled to shut down in 2019. In support of the associated analyses and planning, availability of accurate descriptions of the nuclear waste streams is of paramount importance. The work described in this thesis represents a significant improvement of Nagra's most important radioactive waste characterization methodologies: for spent fuel, NPP decommissioning waste, and reactor waste. The resulting high-fidelity nuclide inventories will decisively support the future waste disposal research and development, as well as serve the NPP decommissioning planning in general. The spent fuel characterization demonstrated is based on Polaris. This approach is compared against the code sequence used in the industry (Studsvik CMS), in order to confirm that, when the necessary fuel data is available, either one of these codes can be used to provide detailed nuclide vector for the spent fuel. Afterwards, possible ways to produce the desired nuclide vectors in the future are identified. For decommissioning waste, every step of the old characterization methodology has been further developed and improved upon. This includes a new, detailed, fully three-dimensional approach to NPP modeling and the development of a new activation sequence (based on the ORIGEN depletion solver), which outputs a highly-resolved component-wise activation distribution and visualization. Additionally, the methodology is extended to apply the method's results for algorithm-optimized packaging concepts and waste volume minimization through decay storage and free release analyses. At the end, recommendations for the future improvements are presented. The decommissioning waste characterization methodology is then expanded to accommodate reactor waste, and the more complex nature of the non-stationary components of this waste stream (such as the control rods). A systematic general approach, based additionally on the component dose rate measurements, allows for a more efficient and flexible characterization of all reactor waste. These developments serve as an important foundation in the ongoing transition from a conservative approach towards a best-estimate plus uncertainty (BEPU) waste characterization, which is vital for accurate safety analysis studies, as well as waste and cost minimization. The results produced by these methodologies will form the basis for the next version of MIRAM (the Swiss Model Inventory of Radioactive Materials), MIRAM2020, which is to be used for the RBG. They will also be used as part of the next Swiss NPP decommissioning cost study, KS2021. At the same time, the obtained results are already being provided to the Swiss NPPs planning their decommissioning, specifically KKM and Beznau (KKB), allowing detailed component-wise segmentation strategies and packaging concepts decision making, finally leading to notable cost reduction for the utilities.

EPFL2019

Currently, most Spent Nuclear Fuel (SNF) is kept safely in storage either at on-site facilities or at centralized interim storage sites. Moreover, many countries face delays in implementing their waste management programmes for SNF and high-level waste disposal. As storage pools approach their capacity, an increasing demand for interim dry-storage solutions is foreseen in the near future and, is associated with emerging regulatory issues. Storage prolongation in interim facilities for time periods far beyond than what originally envisaged, requires license renewal of the facility, as well as of transport and storage casks. At the same time, the safety of SNF operation activities (transportation and handling) at surface encapsulation facilities must be ensured. In order to evaluate the SNF response under normal or accident scenarios, in different stages of the back-end of the nuclear fuel cycle, it is essential to assess the mechanical properties of fuel rods, as well as their evolution during and after irradiation. Several previous studies have identified mechanisms that could affect such properties during dry-storage. However, data generated from experiments with irradiated SNF are extremely rare. This PhD complements previous studies by generating new data for the structural performance of SNF, in areas where technical gaps have been identified. In particular, the mechanical properties of SNF as function of burnup have been investigated both experimentally and numerically, under quasi-static and dynamic bending loads. The experimental activities were conducted at the hot-cell facilities of JRC-Karlsruhe and include three-point bending and gravitational impact tests performed at room temperature on pressurized samples from commercially used PWR UO2 rods. Their dynamic response was studied by successfully applying a new Image Analysis methodology developed in this work. Preliminary tests were conducted on surrogate rodlets consisting of fresh and hydrogenated Zircaloy-4 cladding tubes filled with alumina pellets. The results showed that the sampleâs ductility decreases with increasing hydrogen concentration in the cladding. Five SNF rods, with a wide range of burnup, were selected for the tests on irradiated fuel. High burnup samples showed higher toughness due to irradiation damage on the cladding, and less ductile behavior. Overall, strain rate had marginal influence on ductility for high-burnup samples, whereas for low burnup ones, ductility decreased significantly with strain rate. The flexural mechanical properties of the claddings as function of burnup were derived from the 3-point bending tests using the Euler-Bernoulli beam theory for beam elements with hollow-circular profile. A series of post-test examinations concerning the rod failure processes showed that fuel release in case of rod fracture is very limited, amounting to less than the mass of one pellet. Finite Element Analysis was performed in order to simulate the experiments and the rodsâ bending behavior under the examined load conditions. An extensive sensitivity analysis was performed to minimize modeling uncertainties and the final model was calibrated against experimental data from the 3-point bending tests. Good agreement was observed between numerically predicted and experimentally derived mechanical properties, and the model can be used to simulate the mechanical behavior of SNF rods under different loading configurations.

EPFL2021

Uncertainty Quantification and Representativity Analysis of the LWR-PROTEUS Phase II Experimental Campaign

Daniel Jerôme Siefman

The LWR-PROTEUS Phase II experimental program was conducted at the Proteus research reactor at the Paul Scherrer Institute (PSI) in the early 2000s. One of its purposes was to gain more insight into the reactivity changes caused by fuel burnup and to develop a sense of confidence in modern codes’ ability to predict these changes. The presented project reexamines the experimental campaign using SHARK-X. SHARK-X is a set of Perl-based tools developed at PSI and built around the lattice physics code CASMO-5. It is used to perform sensitivity analysis (SA), uncertainty quantification (UQ), and representativity analysis (RA). This report discusses how SHARK-X was used to quantify the effect of input uncertainties when modeling the LWR-PROTEUS Phase II experiments and to evaluate the representativity of the experiments to a spent fuel pool of the nuclear power plant Gösgen (KKG).

2015

High-Fidelity Determination of Nuclide Inventories for Radioactive Waste Disposal

Valentyn Bykov

EPFL2019

EPFL2021