The fuel test loop consisted an in-pile test section (IPS) and an out-pile system (OPS) is an nuclear fuel irradiation test facility installed in HANARO and its operating temperature and pressure are similar to those of commercial nuclear power plant’s. Penetration pipe connecting the IPS and OPS at the reactor concrete wall is supported by pool-wall pipe support. The existing pool-wall pipe support established in the HANARO have insulations even thought the leak tightness is not ensured. So, the need for an isolation of the insulations from the HANARO cooling water makes the existing pool-wall pipe support newly designed. In this study a structural evaluation for the pool-wall pipe support in accordance with the 2001 ASME B&PV Section III NF is implemented. The most critical primary and secondary stress intensities occur at the modified connection area of the main cooling water pipe and plate ring, but those values are less than the allowable stress. It is concluded that the existing pool-wall pipe support could be modified to a newly designed shape having an isolated insulation from a HANARO cooling water.

In-Pool By-Pass Pipe is a structure which connects the 2 In-Pool Pipes instead of In-Pile Test Section(IPS). It is designed to accommodate the 17.5 MPa and 80 ℃ under the consideration of the FTL pre-operation conditions but the need for high temperature, over 200 ℃, during FTL pre-operation make the additional assessment to be performed. For this study 2 models are used. One is an In-Pool By-Pass Pipe model which affected by HANARO water's elevation, another is an In-Pool By-Pass Pipe Nozzle which has 2 boundary conditions; water and air. After the heat transfer analysis linear stress analysis was performed to achieve Tresca stress. In the region of high stress model's detailed behavior is observed by ASME SectionⅢ NB code. Consequentially it concluded that the model of In-Pool By-Pass Pipe Structure is in reasonable agreement with those code.

The In-Pile Section(IPS) is located inside the reactor pool. It is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. A IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

The small break loss-of-coolant accidents for the HANARO fuel test loop have been predicted by MARS code. Conservative method was used for the prediction of the loss-of-coolant accidents. The maximum peak cladding temperature was calculated as 1286K, which was lower than the design limit temperature (1477K) of nuclear fuels for the HANARO fuel test loop. The maximum peak cladding temperature occurred for the cold leg break in the HANARO pool. The hydrogen generation and oxidation of the fuel cladding were also negligible. Consequently, it is ensured that the emergency cooling water system for the HANARO fuel test loop is appropriate for the small break loss-of-coolant accidents.

FTL(Fuel Test Loop) is a facility that confirms performance of nuclear fuel at a similar irradiation condition with that of nuclear power plant. FTL construction work began on August, 2006 and ended on March, 2007. During Construction, ensuring the worker's safety was the top priority and installation of the FTL without hampering the integrity of the HANARO was the next one. The installation works were done successfully overcoming the difficulties such as on the limited space, on the radiation hazard inside the reactor pool, and finally on the shortening of the shut down period of the HANARO. The Commissioning of the FTL is on due to check the function and the performance of the equipment and the overall system as well. The FTL shall start operation with high burn up test fuels in early 2008 if the commissioning and licensing progress on schedule.

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It is consisted of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS which is localed inside the pool is divided into 3-parts; they are in-pool pipes, IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor collant pressure boundary. The localization of the IVA is achieved by manufacturing through local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique of the instrument lines has been checked for its functionality and yield. A IVA has been manufactured by local technique and will be finally tested under out of the high temperature and high pressure test.

The nuclear fuel cladding temperatures of the HANARO fuel test loop have been calculated by MARS code for the large break loss-of-coolant accidents. Conservative method was used for the analysis of the loss-of-coolant accidents. Consequently, the maximum peak cladding temperature was predicted as 1235K, which was lower than the design limit temperature (1477K) of nuclear fuels for the HANARO fuel test loop. This means that the cooling capability of the emergency cooling water system for the HANARO fuel test loop is sufficient for the large break loss-of-coolant accidents.

The conservative method on the analysis of loss-of-coolant accidents for the HANARO fuel test loop was established based on the guide of evaluation method for the emergency core cooling systems of pressurized light water reactors. The evaluation models, the Moody model for discharge rate calculation and the Baker-Just model for water-metal reaction calculation, were used. In order to calculate conservative peak cladding temperatures for accidents the multipliers to the correlations of heat transfer coefficients in the MARS were also introduced. Consequently it is found that the maximum peak cladding temperature predicted by using the conservative method is sufficiently greater than that calculated by using the best-estimated models.