Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks (original) (raw)
Related papers
A study of the probabilistic risk assessment to the dry storage system of spent nuclear fuel
International Journal of Pressure Vessels and Piping, 2010
Due to the large power supply in the energy market since 1960s, the nuclear power planets have been consistently constructed throughout the world in order to maintain and supply sufficient fundamental power generation. Up to now, most of the planets have been operated to a point where the spent fuel pool has reached its design capacity volume. To prevent the plant from shutdown due to the spent fuel pool exceeding the design capacity, the dry cask storage can provides a solution for both the spent fuel pool capacity and the mid-term storage method for the spent fuel bundles at nuclear power planet.
Reducing the Hazards from Stored Spent Power-Reactor Fuel in the United States
Because of the unavailability of off-site storage for spent power-reactor fuel, the NRC has allowed high-density storage of spent fuel in pools originally designed to hold much smaller inventories. As a result, virtually all U.S. spent-fuel pools have been re-racked to hold spent-fuel assemblies at densities that approach those in reactor cores. In order to prevent the spent fuel from going critical, the fuel assemblies are partitioned off from each other in metal boxes whose walls contain neutron-absorbing boron. It has been known for more than two decades that, in case of a loss of water in the pool, convective air cooling would be relatively ineffective in such a "dense-packed" pool. Spent fuel recently discharged from a reactor could heat up relatively rapidly to temperatures at which the zircaloy fuel cladding could catch fire and the fuel's volatile fission products, 2 Alvarez et al.
Commercial nuclear power plants around the world harness nuclear fission to produce electricity. At each plant, a fission reactor receives fresh nuclear fuel and discharges spent nuclear fuel (SNF). Although the SNF is “spent”, it contains a large amount of radioactive material. Some of that material could be released to the environment by an accident or an attack, causing harm to humans by exposing them to ionizing radiation. The potential for such harm is the “radiological risk” associated with SNF. Independent assessment of this risk could help societies to manage the risk. This report is designed as a handbook that could be used to support such independent assessment. The report has two main parts. The first part provides introductory material, and the second part sets forth a seven-step approach to assessing SNF radiological risk.
Calculation of dose rates due to loss‑of‑coolant accident in open‑pool spent‑fuel storage
Radiation protection and environment, 2018
The objective of the spent‑fuel storage pool is shielding the worker and public from radiation emitted by radioactive decay in the spent fuel and providing a barrier for any radioactive release. In open‑pool multipurpose reactor, the spent‑fuel storage pool is connected to the main pool through the transfer channel. It was prepared to store 528 spent‑fuel elements distributed in two racks that constructed one above the other. The loss‑of‑coolant accident (LOCA) in spent‑fuel storage pool could result in rising of the radiation dose in the reactor building as the water level in the pool falls. The value of the radiation dose rate depends on the height of the water level above the spent fuels, and the number of spent‑fuel elements stored in the storage pool during LOCA. The dose rate calculations were carried out starting from the minimum height which the water level could drop above the spent‑fuel storage racks. The calculations were carried for two cases as follows: the full capacity of both racks and the full capacity of the lower rack only. Monte Carlo N‑Particle Transport MCNP5 code was used to calculate the radiation dose rate above the storage pool and in the control room. The results show that the dose rate in the control room would be lower than the permissible limit when the water level height was 270 and 140 cm for the two cases, respectively. The dose rate above the storage pool would be lower than the permissible limit when the water height above the racks is higher than 385 cm in the first case, and 290 cm for the second case.
Radiological effect of cask‑drop accident in open‑pool spent fuel storage
Radiation Protection and Environment, 2019
This study investigates the radiological risk that may occur during transport of cobalt device after irradiation in open‑pool‑type reactor from the reactor core to the cobalt cell. The cobalt transport process depends on using heavy shielded cask of 3500 kg weight under the water surface of the spent fuel storage pool which may cause a load drop accident. The load drop accident in the spent fuel storage pool would result in damage of 48 spent fuel elements (FE) maximally. Conservative evaluation for the amount of fission products release from the damage of the spent FE was considered in this study by assuming that all spent FE was recently stored at the time of accident. Consequently, the resulting radiation dose distribution was calculated around the reactor building depending on the meteorological data of the reactor site. GENII‑2 code was used to estimate the individual effective dose distribution around the reactor. The result shows that the receptor who located at 1500 m from the reactor building in the south‑east direction will receive the maximum individual effective dose of 0.11 Sv.
Long-term storage of nuclear fuel in spent fuel casks
18TH CONFERENCE OF POWER SYSTEM ENGINEERING, THERMODYNAMICS AND FLUID MECHANICS
SKODA JS has designed a new final disposal cask for long-term storage of nuclear fuel from a VVER-440 reactor. The aim of the research is to investigate the influence of enrichment, burn-up of spent fuel and water introduced during an accident on neutron production in spent fuel. Also, the aim is to investigate the influence of the fuel assemblies with spent fuel on each other. The research is based on the implementation of a new methodology of radiation safety analysis. A radiation field model was created in the initial research stage. A simplified model of 60 fuel assemblies in a square lattice was created. The FMM (Fission Matrix Method) is used for the calculation. Monaco and KENO-6 Monte Carlo neutron transport codes were used for the simulation. The codes are part of the SCALE code package developed at Oak Ridge National Laboratory. The most difficult issue is to meet the safety limit in the criticality safety analysis during the condition of optimum moderation, when moderation is maximum. The most important parameter is the distance between the fuel assemblies. Also, during an accident underground water can be introduced between the fuel assemblies inside the spent fuel cask, and this is why the influence of the water density was investigated, and was found to be very high.
Dose Rate Profile Inside The Spent Fuel Storage Pool in Case of Full Capacity Storage
Journal of Nuclear Physics, Material Sciences, Radiation and Applications, 2020
This study aims to evaluate the radiation dose rate distribution inside temporary spent fuel open-pool storage. The storage pool is connected to the main pool via transfer channel to facilitate transporting the spent fuel under water that avoiding radiation dose rising in the working area in the reactor. The storage pool was prepared to store 800 spent fuel elements that considering the maximum capacity of storage. The spent fuel elements in the storage pool have different decay times depending on the times of extraction from the core. Assuming conservatively, that the spent fuels of the 5-years decay time would be stored in the lower rack and the spent fuels, of decay time ranged between 10 days and 5 years, would be stored in the upper rack. The dose rate was profiled in the region above the upper rack using SCALE/MAVRIC code applying adjoint flux calculation as a variance reduction technique. The results show that the dose rate values in the region above the pool surface would be lower than the permissible limits.
2001
The modulated code system SCALE is used to perform a standardized shielding analysis for any facility containing spent fuel: handling devices, transport cask, intermediate and final storage facility. The neutron and gamma sources as well as the dose rates can be obtained using either discrete-ordinates or Monte Carlo methods. The shielding analysis control modules (SAS1, SAS2H and SAS4) provide a general procedure for cross-section preparation, fuel depletion/decay calculation and general onedimensional or multi-dimensional shielding analysis. The module SAS4 used in the analysis presented in this paper, is a three-dimensional Monte Carlo shielding analysis module, which uses an automated biasing procedure specialized for a nuclear fuel transport or storage container. The Spent Fuel Interim Storage Facility in our country is projected to be a parallelepiped concrete monolithic module, consisting of an external reinforced concrete structure with vertical storage cylinders (pits) arra...