• Journal of Radiation Industry
• /
• v.17 no.3
• /
• pp.283-292
• /
• 2023

Spent nuclear fuel corresponds to high-level radioactive waste that has high decay heat and radioactivity. Accordingly, Spent nuclear fuel withdrawn from the reactor core is primarily stored and managed in a spent nuclear fuel pool in the nuclear power plant to reduce decay heat and radioactivity. In Korea, most nuclear power plant store all spent nuclear fuel in a spent nuclear fuel pool. For wet storage, there are no defense in depth different with reactor core. The study related to spent nuclear fuel pool accident should be carried out to ensure safety. Therefore, it is necessary to analyze previous study cases related to safety of spent nuclear fuel pool and accident cases to build foundational knowledge. The Objective of this study is to analyze study cases of safety assessment and cases for spent nuclear fuel pool accident. For analyzing study cases of safety assessment, possible phenomena when spent nuclear fuel pool accident occurring identified, Subsequently, study cases for safety assessment about each phenomena were investigated, and materials & methods and results for each study are analyzed. For analyzing cases for spent nuclear fuel pool accident, we analyzed accident cases caused by loss of cooling and loss of coolant in spent nuclear fuel pool. Subsequently, causes and change of water level and temperature by each accident case are analyzed. As a result of the analysis on study cases of spent nuclear fuel pool accident, the results of the study conducted by each research institute were vary depending on the computer code, materials & methods of experiment and major assumptions used in the study. As a result of analyzing cases for spent nuclear fuel pool accident, it was found that accident cases for loss of cooling is more than cases for loss of coolant accident. Even though the types of accident in spent nuclear fuel pool were similar, the specific causes were different by each accident case. All the accident cases analyzed did not lead to severe accidents, such as nuclear fuel being exposed to the air. The result of this study will be used as fundamental data for study on spent nuclear fuel pool accident that will be conducted in the future.

• Journal of the Korean Society of Systems Engineering
• /
• v.17 no.2
• /
• pp.82-90
• /
• 2021

The spent fuel pool (SFP) of a nuclear power plant functions to store the spent fuel. The spent fuel pool is designed to properly remove the decay heat generated from the spent fuel. If the cooling function is lost and proper operator action is not taken, the spent fuel in the storage pool can be damaged. Probabilistic safety assessment (PSA) is a safety evaluation method that can evaluate the risk of a large and complex system. So far, the probabilistic safety assessment of nuclear power plants has been mainly performed on the reactor. This study defined the requirements and the functional architecture for the probabilistic safety assessment of the spent fuel pool (SFP-PSA) by applying the systems engineering process. And, a systematic and efficient methodology was defined according to the architecture.

• Nuclear Engineering and Technology
• /
• v.55 no.8
• /
• pp.3102-3113
• /
• 2023

Following the Fukushima nuclear disaster, the simulation of accidents in the spent fuel pool has become more noticeable. Despite the low amount of decay heat power, the consequences of the accidents in a spent fuel pool (SFP) can be severe due to the high content of long-lived radionuclides and lack of protection by the pressure vessel. In this study, the loss-of-cooling accident (LOFA) for the VVER-1000/V446 spent fuel pool is simulated by employing RELAP5 and MELCOR 1.8.6 as the best estimate and severe accident analysis codes, respectively. For two cases with different total power levels, decay heat of spent fuels is calculated by ORIGEN-II code. For modeling SFP of a VVER-1000, a qualified nodalizations are considered in both codes. During LOFA in SFP, the key sequences such as heating up of the pool water, boiling and reducing the water level, uncovering the spent fuels, increasing the temperature of the spent fuels, starting oxidation process (generating Hydrogen and extra power), the onset of fuel melting, and finally releasing radionuclides are studied for both cases. The obtained results show a reasonable consistency between the RELAP5 and MELCOR codes, especially before starting the oxidation process.

• Journal of Radiation Protection and Research
• /
• v.36 no.4
• /
• pp.174-182
• /
• 2011

With the reactor operation conditions - 4.3 wt% $^{235}U$ initial enrichment, burn-up 55,000 MWd/MTU, average power 34 MW/MTU for three periods burned time for 539.2 days per period and cooling time for 100 hours after shut down, to set up the condition to determine the minimum height (depth) of spent fuel storage pool to shut off the radiation out of the spent fuel storage pool and to store spent fuels safely, the dose rate on the specific position directed to the surface of spent fuel storage pool induced by the neutron and gamma-ray from spent fuels are evaluated. The length of spent fuel is 381 cm, and as the result of evaluation on each position from the top of spent fuel to the surface of spent fuel storage pool, it is difficult for neutrons from spent fuels to pass through the water layer of maximum 219 cm (600 cm from the floor of spent fuel storage pool) and 419 cm (800 cm from the floor of spent fuel storage pool) for gamma-ray. Therefore, neutron and gamma-ray from spent fuels can pass through below 419 cm (800 cm from the floor) water layer directed to the surface of spent fuel storage pool.

• Nuclear Engineering and Technology
• /
• v.50 no.5
• /
• pp.709-716
• /
• 2018

A criticality safety analysis was performed for the APR-1400 spent fuel pool region-II to ensure the safe storage of spent fuel, with credit taken for depletion and in-rack neutron absorbers (Metamic panels). PLUS7 fuel assembly was modeled using TRITON-NEWT of SCALE-6.1. The burnup-dependent cross-section library was generated under limiting core-operating conditions with 5%-w U-235 initial enrichment. MCNP5 was used to evaluate the neutron multiplication factor in an infinite array of rack cells with the axially nonuniformly burnt PLUS7 assemblies under normal, abnormal, and accident conditions; including all biases and uncertainties. The main purpose of this study is to investigate reactivity variations due to the critical depletion and reactor operation parameters. The approach, assumptions, and modeling methods were verified by analyzing the contents of the most important fissile and the associated reactivity effects. The Nuclear Regulatory Commission (NRC) guidance on k-eff being less than 1.0 for spent fuel pools filled with unborated water was the main criterion used in this study. It was found that assemblies with 49.0 GWd/MTU and 5.0 w/o U-235 initial enrichment loaded in Region-II satisfy this criterion. Moreover, it was found that the end effect resulted in a positive bias, thus ensuring its consideration.

• Nuclear Engineering and Technology
• /
• v.53 no.4
• /
• pp.1127-1133
• /
• 2021

A Spent Fuel Pool (SFP) is designed to store spent fuel assemblies in the pool. And, a SFP cooling and cleanup system cools the SFP coolant through a heat exchanger which exchanges heat with component cooling water. If the cooling system fails or interfacing pipe (e.g., suction or discharge pipe) breaks, the cooling function may be lost, probably leading to fuel damage. In order to prevent such an incident, it is required to properly cool the spent fuel assemblies in the SFP by either recovering the cooling system or injecting water into the SFP. Probabilistic safety assessment (PSA) is a good tool to assess the SFP risk when an initiating event for the SFP occurs. Since PSA has been focused on reactor-side so far, it is required to study on the framework of PSA approach for SFP and identify the key factors in terms of fuel damage frequency (FDF) through a case study. In this study, therefore, a case study of SFP-PSA on the basis of design information of APR-1400 has been conducted quantitatively, and several sensitivity analyses have been conducted to understand the impact of the key factors on FDF.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.14 no.2
• /
• pp.149-156
• /
• 2016

The spent fuel stored in the pool is vulnerable to external impacts, since the severe reactor conditions degrade the structural integrity of the fuel. Therefore an accident during shipping and handling should be considered. In an extreme case, the fuel assembly drop can be happened accidentally during handling the nuclear fuel in the spent fuel pool. The rod failure during such drop accident can be evaluated by calculating the impact force acting on the fuel assembly at the bottom of the spent fuel pool. The impact force can be evaluated with the impact velocity at the bottom of the spent fuel pool. Since fuel rods occupies most of weight and volume of a nuclear fuel assembly, the information of the rods are important to estimate the hydraulic resistance force. In this study, the hydraulic force acting on the $3{\times}3$ short rod bundle model during the drop accident is calculated, and the result is verified by comparing the numerical simulations. The methodology suggested by this study is expected to be useful for evaluating the integrity of the spent fuel.

• Nuclear Engineering and Technology
• /
• v.55 no.6
• /
• pp.2088-2095
• /
• 2023

The safety of nuclear installations is largely determined by the tightness of fuel elements cladding. As the Fukushima nuclear accident showed, the main task in case of loss of power supply is to ensure reliable removal of residual heat release from spent fuel pool (SFP) with irradiated fuel assemblies (IFAs). The paper presents the results of calculated-experimental studies and thermal-hydraulic modeling of temperature storage modes of IFAs in SFP. Experimental studies of SFP's temperature regime and calculated evaluation of residual heat removal due to the thermal conductivity of building structures surrounding the SFP were performed. To ensure the safe operation of research reactors, it's necessary to know the IFA's residual heat power (RHP) in the reactor and SFP, which is determined depending on the operating time of fuel assemblies (FAs) and the IFAs calculated holding time. The FAs operating time depends on the reactor energy output. The IFAs calculated holding time is determined by the fuel burnup, U-235 mass in the fuel, and reactor utilization factor. The IFAs fuel burnup was calculated using the MCU-PTR program. Also presented are the RHP's calculation results using some of the empirical dependencies. The concept of a passive heat removal system (PHRS) based on thermosyphon's operating principle was proposed.

• Nuclear Engineering and Technology
• /
• v.52 no.3
• /
• pp.654-659
• /
• 2020

The reliability of the spent fuel pool instrument is very important for the security of nuclear power plant, especially during the earthquake. The effect of the fluid force on the vibration characteristics of the flexible tube of the spent fuel pool instrument needs comprehensive analysis. In this paper, based on the potential flow theory, the hydrodynamic pressures acting on the flexible tube were obtained. A mathematical model of a flexible tube was constructed to obtain the dynamic response considering the effects of seismic load and fluid force, and a computer code was written. Based on the mathematical model and computer code, the maximum stresses of the flexible tube in both safe shutdown earthquake and operating basis earthquake events on the spent fuel pool with three typical water levels were calculated, respectively. The results show that the fluid force has an obvious effect on the stress and strain of the flexible tube in both safe shutdown earthquake and operating basis earthquake events.

• Nuclear Engineering and Technology
• /
• v.51 no.3
• /
• pp.884-893
• /
• 2019

Spent fuel racks are steel structures used in the storage of the spent fuel removed from the nuclear power reactor. Rack units are submerged in the depths of the spent fuel pool to keep the fuel cool. Their free-standing design isolates their bases from the pool floor reducing structural stresses in case of seismic event. However, these singular features complicate their seismic analysis which involves a transient dynamic response with geometrical nonlinearities and fluid-structure interactions. An accurate estimation of the response is essential to achieve a safe pool layout and a reliable structural design. An analysis methodology based on the hydrodynamic mass concept and implicit integration algorithms was developed ad-hoc, but some dispersion of results still remains. In order to validate the analysis methodology, vibration tests are carried out on a reduced scale mock-up of a 2-rack system. The two rack mockups are submerged in free-standing conditions inside a rigid pool tank loaded with fake fuel assemblies and subjected to accelerations on a unidirectional shaking table. This article compares the experimental data with the numerical outputs of a finite element model built in ANSYS Mechanical. The in-phase motion of both units is highlighted and the water coupling effect is detailed. Results show a good agreement validating the methodology.