Loss-of-coolant Accident Tests Research Articles

Comparison of accelerated and non-accelerated post-LOCA simulation

Cables important to safety used in nuclear power plants are required to withstand environmental, operational and system conditions to which they are exposed during their design life. This may include harsh conditions resulting from a design basis event (DBE), such as a Loss-of-Coolant Accident (LOCA). The testing procedure includes accelerated thermal and ionizing radiation ageing which brings the tested samples to their end-of-life condition. Accelerated pre-ageing is then followed by accident simulation (LOCA test). The latter part of accident simulation (post-LOCA), which may take up to 1 year, is commonly accelerated based on the Arrhenius approach, however the adequacy of this approach remains a topic of ongoing discussion, lacking a clear scientific or engineering consensus. While numerous works have focused on testing various conditions during accidental and post-accidental conditions in nuclear power plants, there is notable absence of direct assessment of acceleration effects.The aim of this work was to assess the effects of accelerating post-accident conditions on electrical cables using non-destructive electrical measurements. The cables were subjected to thermal and radiation pre-ageing, LOCA test, and then were divided into two groups for post-LOCA tests: one group was tested for one year in non-accelerated post-accident conditions, while the other was accelerated using the Arrhenius approach. Diagnostic methods, including insulation resistance, capacitance, voltage withstand tests, and high-frequency parameters such as attenuation, characteristic impedance, and relative permittivity, were employed for comparing sample conditions. Results from both accelerated and non-accelerated post-LOCA simulations were consistent, indicating that the acceleration was adequate.

Thermal-Hydraulic and Engineering Evaluations of New LOCA Testing Methods in TREAT

Design evaluations and thermal-hydraulic predictions were undertaken to compare three candidate options for loss-of-coolant accident (LOCA)–testing capabilities at the Transient Reactor Test facility (TREAT). These options included a small water capsule, a large natural circulation capsule, and a forced-convection water loop. All options were found to create reasonable approximations of radial temperature gradients in fuel specimens prior to blowdown to a secondary tank. Given its minimalist form factor, the small water capsule was predicted to perform remarkably well in mimicking the thermal response of a 20-pellet fuel rodlet during a postulated LOCA event, while sensitivity studies showed versatility in adjusting conditions for specific test phenomena priorities. The large natural circulation capsule predictions showed representative capsule pressure evolution but indicated it would likely struggle to capture the desired fuel temperature response in the blowdown transition, primarily due to nonrepresentative cooling as its tall water column was discharged. These results were compared to previous water loop evaluations that had shown a capability for representing ideal test conditions, especially for longer rodlet test specimens in small bundles, while being achievable in a self-contained form factor to facilitate test logistics. Considering the broader cost and capability spectrum needed to support fuel safety research, these efforts resulted in a recommendation to pursue the small water capsule for near-term data needs specific to high-burnup fuel fragmentation, while continuing detailed development of the water loop to address ultimate data needs.

Integral LOCA fragmentation test on high-burnup fuel

Increasing the fuel burnup limit in light water reactors is sought to enhance fuel cycle economics and requires establishing a technical basis. Experimental observations of severe fuel fragmentation under loss-of-coolant-accident conditions at Halden and Studsvik had raised the need for additional considerations during the development of this technical basis. These test data suggested that the burnup threshold for high burnup fuel fragmentation may be influenced by pre-transient power. Additional loss-of-coolant-accident test data is therefore valuable to complement these tests and enhance the current state of understating. Oak Ridge National Lab has developed the Severe Accident Test Station capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident and beyond design basis accident scenarios. Severe Accident Test Station provides various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident testing, defueled oxidation quench testing, and high-temperature beyond design basis accident testing. Severe Accident Test Station has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory at Oak Ridge National Lab. Furthermore, descriptions of the in-cell re-fabrication capabilities and assembly of the loss-of-coolant-accident test train are provided. Installation of the Severe Accident Test Station system and in-cell re-fabrication restores United States capability to examine postulated and extended loss-of-coolant-accident conditions on spent fuel and cladding and provides a platform for evaluating advanced fuel and accident-tolerant fuel cladding concepts. Lastly, three in-cell integral loss-of-coolant-accident test were performed in the Severe Accident Test Station and compared to the Nuclear Regulatory Commission sponsored loss-of-coolant-accident test as well as the original loss-of-coolant-accident test performed at Argonne National Lab. The results of these tests as well as all publicly available integral loss-of-coolant-accident test were used to validate a threshold for high burnup fuel fragmentation.

Characterization of the relocated and dispersed fuel in the Halden reactor project LOCA tests based on gamma scan data

The on-going Loss-of-Coolant Accident (LOCA) test program at the OECD Halden Reactor Project (HRP) conducts in-house gamma scanning as standard post-irradiation examination (PIE) procedure on Light Water Reactor (LWR) fuel rods. One of the primary objectives of the program is to investigate fuel relocation into the balloon region and fuel dispersal through the cladding rupture opening after burst. A simple model called Gamma Transport Model was formulated for the purpose of interpretation of fuel relocation based on the gamma scan data. Fuel relocation may have a strong effect on the linear heat generation rate at the balloon due to, firstly, increase in linear fuel density, and secondly due to differences in burn-up and local heat generation rate at the periphery and bulk of the pellet. For this analysis, a pair of short-lived isotopes with very different fission product yields for 235U and 239Pu is selected from the gamma scan spectrum. The intention is to use the difference in the ratio of their concentrations in the balloon region to qualitatively make conclusion on the fuel relocation. As a separate outcome, the same analysis can be applied to the dispersed fuel region and to draw conclusion on its origin (pellet rim or bulk). The Gamma Transport Model is validated against a special (non-destructive) case from the Halden LOCA test program and then applied for the analysis of selected tests. In addition, a methodology is presented for estimation of the amount of dispersed fuel from the LOCA tests based on the gamma scan data. Currently, at Halden there is no possibility to measure the dispersed fuel and hence an alternative is needed. Such information can be used for code validation and for inferring additional information based on other test parameters such as cladding rupture opening.

Neutron radiography and tomography applied to fuel degradation during ramp tests and loss of coolant accident tests in a research reactor

Neutron radiography (NR) is performed at the Institute for Energy Technology (IFE) in Norway since the late 1970s. The application of the non-destructive method was to acquire post-irradiation examination (PIE) data (e.g. fuel integrity and hydrogen up-take in cladding) from safety and integrity tests of nuclear fuels performed under the Organization for Economic Co-operation and Development (OECD) Halden Reactor Project (HRP). The method was later applied under re-fabrication and instrumentation operations of experimental nuclear fuel rods prior to testing in Halden Boiling Water Reactor (HBWR), and for a variety of PIE projects, e.g. reactor power ramp testing, PCI failure detection and fuel degradation experiments. Neutron radiography has also proved to be a very useful tool for examination of nuclear fuels irradiated in the Loss-of-Coolant Accident (LOCA) experimental series initiated in the early 2000s. Neutron tomography data is acquired while an increased international focus arose on fuel fragmentation, fuel relocation and fuel dispersal processes that occur during the LOCA events for high burn-up nuclear fuels. Hydrogen up-take of the fuel cladding, fuel pellet-clad bonding condition, fuel fragmentation, particle size distributions, and other features obtained from neutron tomography data are quite relevant for reactor core safety impact study of LOCA events simulated in the HBWR. Neutron tomography studies of LOCA tested fuel were done in cooperation with the SCK·CEN institute in Mol, Belgium, and the University of Antwerp in Belgium. It’s interesting to observe that the image reconstruction results obtained from the SART method are quite good regarding the relatively few sample rotations utilized under acquisition of neutron radiography projections in the tomography studies of the LOCA examination.

Major Achievements and Prospect of the ATLAS Integral Effect Tests

A large-scale thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation), has been operated by KAERI. The reference plant of ATLAS is the APR1400 (Advanced Power Reactor, 1400 MWe). Since 2007, an extensive series of experimental works were successfully carried out, including large break loss of coolant accident tests, small break loss of coolant accident tests at various break locations, steam generator tube rupture tests, feed line break tests, and steam line break tests. These tests contributed toward an understanding of the unique thermal-hydraulic behavior, resolving the safety-related concerns and providing validation data for evaluation of the safety analysis codes and methodology for the advanced pressurized water reactor, APR1400. Major discoveries and lessons found in the past integral effect tests are summarized in this paper. As the demand for integral effect tests is on the rise due to the active national nuclear R&D program in Korea, the future prospects of the application of the ATLAS facility are also discussed.

Qualification of Cables to IEEE Standards 323-1974 and 383-1974

Wire and Cable manufacturers generally qualify products for class IE application by envelope type testing to user specifications and environmental conditions recommended by IEEE Standards 323-1974 and 383-1974. The General Electric Wire and Cable Business Department recently completed two such qualification programs. Cable constructions tested were 600V control cables and 600V, 2KV, and 15KV power cables insulated with flame resistant mineral filled crosslinked polyethylene. The 15KV samples included taped field splices. To determine the ability of samples to withstand a loss of coolant accident (LOCA) both early in and at the end of design life, both thermally aged and unaged cables were irradiated and subjected to a 100 day plus 10 day margin simulated LOCA test. Samples were heat aged to a time-temperature condition in excess of 40 years of service life. The aging criterion was determined from experimental data fitted to an Arrhenius curve by the Regression Method of IEEE Standard 101-72. All test specimens were irradiated using a Cobalt 60 source to a total integrated dose of 220 Mrad. The simulated LOCA conditions of one test program followed the combined requirements for Boiling Water Reactors and Pressurized Water Reactors. In the second test program, the steam pressure-temperature profile included a simulated main steam line break. Test specimens were wrapped on grounded mandrels and were electrically loaded throughout the simulated LOCA tests. After completion of environmental testing, samples were subjected to the IEEE 383 simulated post-LOCA test.

Simulated and Simultaneous Loss-of-Coolant Accident Testing of Protective Coatings for the Nuclear Industry

An apparatus was designed and constructed to test the performance of coatings under conditions of high-pressure steam and radiation that might exist under a loss-of-coolant accident (LOCA). Results from the ''simultaneous'' exposure of coatings to high-pressure steam and radiation are compared to results obtained from the conventional ''simulated'' test procedures. Coating systems were selected that had a history of performing well under simulated LOCA conditions and included zinc-based, epoxy, and phenolic primers with phenolic and modified phenolic topcoats. Coatings were exposed to $sup 60$Co radiation doses in the range of 10$sup 8$ to 10$sup 9$ rad. The study showed that the conventional simulated LOCA conditions were more severe on the coatings than those tested under the simultaneous exposure to high-pressure steam and $sup 60$Co radiation. It was concluded that coatings that satisfactorily passed the simulated LOCA tests will also pass the simultaneous LOCA tests.