• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.167-174
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• 2011

A pressurizer in a small integral type pressurized water reactor is located inside the upper region of the reactor vessel, and uses a space between the upper head of the reactor vessel and the upper region of the upper guide structure which is partially filled with fluid depending on the operating power. This new design requires a comprehensive investigation of vibration characteristics. This study investigates the modal characteristics of a pressurizer which uses a simplified cylindrical shell model, focusing on how having fluid in the shell affects vibration and response characteristics. In addition, an analysis of sloshing is performed and the response characteristics are addressed.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.63-77
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• 1994

The objective of this work is to investigate fluid mixing phenomena related to pressurized thermal shock(PTS) in a pressurized water reactor(PWR) vessel downcomer during transient cooldown with direct vessel injection(DVI) using test models. The test model designs were based on ABB Combustion Engineering(C-E) System 80＋ reactor geometry. A cold leg small break loss-of-coolant accident(LOCA) md a main steam line teak were selected as the potential PTS events for the C-E System 80＋. This work consist of two parts. The first part provides the visualization tests of the fluid mixing between DVI fluid and existing coolant in the downcomer region, and the second part is to compare the results of thermal mixing tests with DVI in the other test model. Row visualization tests with DVI have clarified the physical interaction between DVI fluid and primary coolant during transient cooldown. A significant temperature drop was observed in the downcomer during the tests of a small break LOCA Measured transient temperature profiles agree well with the predictions by the REMIX code for a small break LOCA and with the calculations by the COMMIX-1B code for a steam line break event.

• Journal of the Korean Society of Systems Engineering
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• v.13 no.1
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• pp.25-31
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• 2017

This paper aims to evaluate the conservatism in the design of APR1400 (Advanced Pressurized water Reactor 1400 designed by KHNP) reactor internals component, the LSS (Lower Support Structure). Re-engineering of the LSS is done based on the system design condition data and applicable ASME code that was used for the original APR1400 design. Systems engineering approach is applied to design the LSS of APR1400 without refering APR1400 LSS dimensional parameters and tries to verify important design parameters of APR1400 LSS as well as the validity of the re-engineering design process as independent verification method of reactor component design. Systems engineering approach applied in this study following V-model approach. The re-engineered LSS design showed more than enough conservatism for static loading case. The maximum deflection of LSS is under 1mm (calculated value is 0.25mm) from 4000 mm diameter of LSS. Hence the deflection can be ignored in other reactor internals for structural integrity assessment. Especially the effect of LSS deflection on fuel assembly can be minimized and which is one of the main requirements of LSS design. It also showed that the maximum stress intensity is 2.36MPa for the allowable stress intensity of 60.1 MPa. The stress resulted from the static load is also very small compared to the maximum allowable stress intensity, hence there is more than enough conservatism in the LSS design.

• International Journal of Safety
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• v.3 no.1
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• pp.53-58
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• 2004

SMART-P is a promising advanced small and medium category nuclear power reactor. It is an integral type reactor with a sensible mixture of new innovative design features and proven technologies aimed at achieving a highly enhanced safety and improved economics. The enhancement of the safety and reliability is realized by incorporating inherent safety improving features and reliable passive safety systems. The improvement in the economics is achieved through a system simplification, and component modularization. Preliminary safety analyses on selected limiting accidents confirm that the inherent safety improving design characteristics and the safety system of SMART-P ensure the reactor's safety. SMART-P is an advanced integral pressurized water reactor. The purpose of this study is for the safety analysis of the steam generator module pipe break for the SMART-P. The integrity of the fuel rod is the major criteria of this analysis. As a result of this analysis, the safety of the RCS and the secondary system is guaranteed against the module pipe break of a steam generator of the SMART-P.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3256-3263
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• 2021

The main steam line break accident is an essential initiating event of the pressurized water reactor. In present work, the fuzzy set theory and the signal-based fault detection method has been used to detect the occurrence and diagnosis of the location and break area for the small scale MSLB. The models are validated by the AP1000 accident simulator based on MAAP5. From the test results it can be seen that the proposed approach has a rapid and proper response on accident detection and location diagnosis. The method proposed to evaluate the break area shows good performances for small scale MSLB with the relative deviation within ±3%.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2880-2886
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• 2020

A small leak occurring on the surface of a fuel rod due to damage exposes UO₂ to a steam atmosphere. During this time, fission gas trapped inside the fuel rod leaks out, and the gas leakage can be increased due to UO₂ oxidation. Numerous studies have focused on the steam oxidation and its thermodynamic calculation in UO₂. However, the thermodynamic calculation of the UO₂ oxidation in a pressurized water reactor (PWR) environment has not been studied extensively. Moreover, the kinetics of the oxidation of UO₂ pellet also has not been investigated. Therefore, in this study, the thermodynamics of UO₂ oxidation under steam injection due to a damaged fuel rod in a PWR environment is studied. In addition, the diminishing radius of the UO₂ pellet with time in the PWR environment was calculated through an experiment simulating the initial time of steam injection at the puncture.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• Journal of the Korea Institute of Military Science and Technology
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• v.20 no.4
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• pp.528-535
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• 2017

Fuel cells have been spotlighted in the world for being highly efficient and environmentally friendly. A hydrogen which is the fuel of fuel cell can be obtained from a number of sources. Hydrogen source for operating the polymer electrolyte membrane fuel cell(PEMFC) in the current underwater environment, such as a submarine and unmanned underwater vehicles are currently from the metal hydride cylinder. However, metal hydride has many limitations for using hydrogen carrier, such as large volume, long charging time, limited storage capacity. To solve these problems, we suggest diesel reformer for hydrogen supply source. Diesel fuel has many advantages, such as high hydrogen storage density, easy to transport and also well-infra structure. However, conventional diesel reforming system for PEMFC requires a large volume and complex CO removal system for lowering the CO level to less than 10 ppm. In addition, because the preferential oxidation(PROX) reaction is the strong exothermic reaction, cooling load is required. By changing this PROX reactor to hydrogen separation membrane, the problem from PROX reactor can be solved. This is because hydrogen separation membranes are small and permeable to pure hydrogen. In this study, we conducted the pressurized diesel reforming and water-gas shift reaction experiment for the hydrogen separation membrane application. Then, the hydrogen permeation experiments were performed using a Pd alloy membrane for the reformate gas.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.32 no.2
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• pp.177-185
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• 2008

Low-cycle environmental fatigue tests of cast austenitic stainless steel CF8M at the condition of fatigue strain rate 0.04%/sec were conducted at the pressure and temperature, 15MPa, $315^{\circ}C$ of a operating pressurized water reactor (PWR). The used test rig was limited to install an extensometer at the gauge length of the cylindrical fatigue specimen inside a small autoclave. So the magnet type LVDT#s were used to measure the fatigue displacement at the specimen shoulders inside the high temperature and high pressure water autoclave. However, the displacement and strain measured at the specimen shoulders is different from the one at the gauge length for the geometry and the cyclic strain hardening effect. Displacement of the fatigue specimen gauge length calculated by FEM (finite element method) used to modify the measured displacement and fatigue life at the shoulders. A series of low cycle fatigue life tests in air and PWR conditions simulating the cyclic strain hardening effect verified that the FEM modified fatigue life was well agreed with the simulating test results. The process and method developed in this study for the environmental fatigue test inside the small sized autoclave would be so useful to produce reliable environmental fatigue curves of CF8M stainless steel in pressurized water reactors.