In this study, the structural integrity of the composite rocket motor case of a space launch vehicle was evaluated by conducting compression and bending tests. Two composite rocket motor case specimens with different stacking patterns were prepared for each test, and a dedicated jig was designed and manufactured. The test procedure was developed and applied separately for compression and bending tests. By performing these tests, the composite rocket motor case structural safety was assessed.

The influence of specimen geometry and notch on the hydrogen embrittlement of an SA372 steel for pressure vessels was investigated in this study. A slow strain-rate tensile (SSRT) test after the electrochemical hydrogen charging method was conducted on four types of tensile specimens with different directions, shapes (plate, round), and notches. The plate-type specimen showed a significant decrease in hydrogen embrittlement resistance owing to its large surface-to-volume ratio, compared to the round-type specimen. It is well established that most of the hydrogen distributes over the specimen surface when it is electrochemically charged. For the round-type specimens, the notched specimen showed increased hydrogen susceptibility compared with the unnotched one. A notch causes stress concentration and thus generates lots of dislocations in the locally deformed regions during the SSRT test. The solute hydrogen weakens the interactions between these dislocations by promoting the shielding effect of stress fields, which is called hydrogen-enhanced localized plasticity mechanisms. These results provide crucial insights into the relationship between specimen geometry and hydrogen embrittlement resistance.

The dismantling of the reactor pressure vessel has been carried out at a number of commercial nuclear power plants, including the Zion nuclear power plant in the United States and the Stade nuclear power plant in Germany. The dismantling method for the reactor pressure vessel is either in the air or in the water, depending on the utility. In general, a mechanical cutting method is used when dismantling the reactor pressure vessel in the water. And when dismantling a nuclear reactor pressure vessel in the air, the thermal cutting method is applied. However, there is no case of dismantling commercial nuclear reactor pressure vessel by applying a mechanical method in the air. In this study, when a nuclear reactor pressure vessel is dismantled by applying a mechanical method in the air, the applicability was evaluated by testing it using a demonstration mockup of Kori Unit 1. For the evaluation, the mockup was made in the actual size of Kori Unit 1. Mechanical cutting devices used the band saw and the circular saw. In the test, the cutting of the reactor pressure vessel was performed remotely by reflecting the working conditions of the decommissioning site. The band saw cutting method was applied to vertical cutting, and the circular saw cutting method was applied to horizontal cutting. In order to dismantle one cut-off piece, mockup test was performed according to a series of dismantling processes, it consists of preparatory work, vertical cutting process, horizontal cutting process, packaging process and finishing work. The cutting speed of the band saw is 3–10 mm·min−1, and the cutting speed of the circular saw is 2–4 mm·min−1. As a result of the test, when the mechanical cutting method was applied, as is known, the kerf width was smaller than when the thermal cutting method was applied. The cut surface showed a clean state without drag lines generated during thermal cutting. However, the working time was much slower than when the thermal cutting method was applied.

Cutting reactor pressure vessels (RPV) into acceptable sizes for waste disposal is a key process in dismantling nuclear power plants. In the case of Kori-1, a remote oxyfuel cutting method has been developed by Doosan Heavy Industry & Construction to dismantle RPVs. Cutting radioactive material, such as RPV, generates a large number of fine and ultrafine particles incorporating radioactive isotopes. To minimize radiological exposure of dismantling workers and workplace surface contamination, understanding the characteristics of radioactive aerosols from the cutting process is crucial. However, there is a paucity of knowledge of the by-products of the cutting process. To overcome the limitations, a mock-up RPV cutting experiment was designed and established to investigate the characteristics of fine and ultrafine particles from the remote cutting process of the RPV at the Nuclear Decommissioning Center of Doosan Heavy Industry & Construction. The aerosol measurement system was composed of a cutting system, purification system, sampling system, and measurement device. The cutting system has a shielding tent and oxyfuel cutting torch and remote cutting robot arm. It was designed to prevent fine particle leakage. The shielding tent acts as a cutting chamber and is connected to the purification system. The purification system operates a pressure difference by generating an airflow which delivers aerosols from the cutting system to the purification system. The sampling system was installed at the center of the pipe which connects the shielding tent and purification system and was carefully designed to achieve isokinetic sampling for unbiased sampling. Sampled aerosols were delivered to the measurement device. A high-resolution electrical low-pressure impactor (HR-ELPI+, Dekati) is used to measure the size distribution of inhalable aerosols (Aerodynamic diameter: 6 nm to 10 μm) and to collect size classified aerosols. In this work, the mock-up reactor vessel was cut 3 times to measure the number distribution of fine and ultrafine particles and mass distribution of iron, chromium, nickel, and manganese. The number distribution of aerosols showed the bi-modal distribution; two peaks were positioned at 0.01−0.02 μm and 0.04–0.07 μm respectively. The mass distribution of metal elements showed bi-modal and trimodal distribution. Such results could be criteria for filter selection to be used in the filtration system for the cutting process and fundamental data for internal dose assessment for accidents. Future work includes the investigations relationships between the characteristics of the generated aerosols and physicochemical properties of metal elements.

Thick-walled pressure vessel has been autofrettaged in order to improve the fatigue life of the pressure vessel. The compressive tangential residual stress near the bore of the pressure vessel due the autofrettage process is benefical to the fatigue crack initiation and propagation of the pressure vessel. However, a reverse yielding due to the Bauschinger effect during the unloading process in autofrettage causes the reduction of the compressive residual stress near the bore. In order to evaluate the fatigue crack propagation life of the autofrettaged thick-walled pressure vessel, the Bauschinger effects were considered. Stress intensity factors of the crack at the inside surface of the pressure vessel due to operating pressure loading of 707 MPa and autofrettage loading with different levels of overstrain were calculated by using finite element methods, and used for evaluating fatigue crack propagation lives. Fatigue lives of the pressure vessel with the Bauschinger effects resulted in 45% to 67% reductions in fatigue life compared to those of the pressure vessel with ideal residual stress distributions depending on the autofrettage level.

Application of a very high internal pressure on the thick-walled pressure vessel induces beneficial compressive tangential residual stresses near the bore of the pressure vessel after unloading the internal pressure. However, a reverse yielding due to the Bauschinger effect during the unloading process causes the reduction of the compressive residual stress near the bore. In order to evaluate the autofrettage residual stress distributions of the thick-walled pressure vessel, the Bauschinger effects were considered. Magnitudes of the compressive residual stresses at the bore determined by considering the Bauschinger effects decreased by about 25 percent, compared to the case of linear elastic unloading, i.e., without Bauschinger effects. Measured residual stress distributions agreed fairly well with the calculated distributions considering the Bauschinger effects.

In this study, the effect of carbon equivalent and cooling rate on microstructure and hardness of A516 steels for pressure vessel is investigated. Six kinds of specimens are fabricated by varying carbon equivalent and cooling rate, and their microstructures and hardness levels are analyzed. Specimens with low carbon equivalent consist of ferrite and pearlite. As the cooling rate increases, the size of pearlite decreases slightly. The specimens with high carbon equivalent and rapid cooling rates of 10 and 20 oC/s consist of not only ferrite and pearlite but also bainite structure, such as granular bainite, acicular ferrite, and bainite ferrite. As the cooling rate increases, the volume fractions of bainite structure increase and the effective grain size decreases. The effective grain sizes of granular bainite, acicular ferrite, and bainitic ferrite are ~20, ~5, and ~10 μm, respectively. In the specimens with bainite structure, the volume fractions of acicular ferrite and bainitic ferrite, with small effective grains, increase as cooling rate increases, and so the hardness increases significantly.

압력용기의 내압은 압력용기 설계의 중요한 인자이며 이를 바탕으로 관련 설계기준 및 구조해석결과에 따라 압력용기의 두께 및 직경과 같은 기하학적 형상이 결정된다. 그러나 압력용기 내부에서 폭굉이 일어날 경우 이 폭굉압력을 적절히 고려 하여 압력용기를 설계할 수 있는 설계기준은 미흡한 실정이다. 일반적으로 폭굉이 발생할 경우, 초기 폭굉압력이 용기 벽면에 도달하여 반사하는 반사압력은 초기압력의 2배 이상이라고 알려진다. 그러나 폭굉압력은 구조물의 고유주기보다도 짧은 시간 안에 최대치에 도달한 후 급격하게 감소하는 경향을 보이며, 이 경우 실제 용기벽면이 받게 되는 압력은 반사압력에 비해 매우 작을 수 있다. 따라서 본 연구에서는 이러한 폭굉의 특성을 고려하여 압력용기가 견뎌야 하는 적절한 등가의 폭굉압 력을 산정하는 방법을 제안함으로써 폭굉을 고려한 효율적인 압력용기 설계기준을 제시하고자 하였다.

본 연구는 50대 비만중년여성을 대상으로 순환운동이 건강체력, 혈압 및 혈관탄성도에 미치는 영향을 규명하는데 있었으며, 이를 위해 비만중년여성 24명을 대상으로 운동그룹 12명, 대조군 12명으로 분류하여 8주간 순환운동 프로그램을 실시하였다. 순환운동 프로그램은 유산소 운동과 저항운동으로 구성 되었으며, 회당 40분간, 주3회 실시하였고, 유산소 운동 수행강도는 50-70% HRmax, 저항운동은 1-RM 의 40-60%로 수행되었으며, 순환운동 전·후에 측정한 자료의 그룹 내 차이 비교를 위해 대응표본 T검정, 그룹 간 차이는 운동 전·후의 변화율을 산출하여 독립표본 T검정을 실시하였다. 그 결과 운동그룹은 건강 체력 중 BMI(p<.05)가 유의하게 감소하였고, 심폐지구력(p<.001), 근지구력(p<.001), 유연성(p<.01)이 증가하였으며, 그룹 간 대조군 보다 심폐지구력, 유연성이 유의하게 증가되었다(p<.01). 수축기 혈압과 이완기 혈압은 운동그룹이 유의하게 감소하였으며(p<.01), 이완기 혈압은 그룹 간 대조군 보다 운동그룹이 유 의하게 감소하였다(p<.05). PWV는 운동그룹이 유의하게 감소하였고(p<.05), 대조군이 유의하게 증가하였으며(p<.01), 그룹 간 대조군 보다 운동군이 유의하게 감소하였다(p<.05). 이상의 결과 비만중년여성의 건강증진 및 비만치료를 위한 순환운동은 건강체력, 혈압 및 혈관탄성도에 긍정적인 영향을 미친 것으로 사료된다.

In order to investigate the low-cycle fatigue behavior of Inconel 718 alloy used for pressure vessels, the strain-controlled fatigue test was performed in the room and high temperatures of 550°C. High temperature test was done using an electric furnace attached on the hydraulic fatigue test system. Tensile strength and elastic modulus of the Inconel 718 alloy at the temperature of 550°C decreased by 8% and 10%, respectively, compared to those at the room temperature. Subjected to the repeated cyclic loading under the strain-control, the material exhibited cyclic softening behavior with decreasing yield strength at both room and high temperatures. The low-cycle fatigue properties determined in this research could be effectively used for the fatigue life estimation of high temperature components made of Inconel 718 alloy.

재래식 해수담수화 수평적 압력용기 설계는 후단에 있는 역삼투막에는 농축수에 의해 수질악화 및 생산수저감 등이 동반된다. 이를 해결하기 위해 본 연구에서는 개념적인 중앙 주입식 압력용기가 생산수 단가에 미치는 영향과 역삼투 공정 설계 시 주요 인자들이 성능에 미치는 영향을 조사를 위해 상업화된 역삼 투 프로그램인 ROSA를 이용하여 분석하였다. 그 결과 중앙 주입식 압력용기를 이용할 경우 총괄적인 회수율과 SEC 측면에서 성능이 저하하지만, 막 모듈 당 생산되는 생산수량이 증가하는 것으로 나타났다. 또한 생산수의 수질개선은 2단 설계의 용량 감소로 인하여 건설비 저감으로 연계되는 것으로 조사되었다. 본 연구는 국토교통부 플랜트연구사업의 연구비지원(과제번호 16IFIP-B089908-03)에 의해 수행되었습니다.

본 논문에서는 노심용융사고 시 관통노즐이 제거된 원자로용기 하부헤드의 구조 건전성 평가를 수행하였다. 열응력, 노심용융물의 질량 그리고 내압조건의 해석결과를 고려할 때, 하부헤드의 열응력에 의한 영향이 가장 크게 나타났다. 손상 가능성은 파손기준에 따라 평가하였으며, 등가소성변형률이 임계변형률 파손기준보다 낮은 수준으로 평가되었다. 열-구조물 연성해석 결과 하부헤드의 두께 중간층에서 항복강도보다 낮은 응력이 발생한 탄성영역 구간을 확인하였다. 내압이 커지면서 탄성영역 범위가 점차 좁아지면서 탄성영역이 내벽으로 이동하는 결과를 확인하였고, 노심용융사고 시 구조적 건전성을 만족하는 것으로 평가되었다.

Fuel Test Loop(FTL) is a facility which could conduct a fuel irradiation test at HANARO (High-flux Advanced Neutron Application Reactor). FTL simulates commercial NPP’s operating conditions such as the pressure, temperature and neutron flux levels to conduct the irradiation and thermo- hydraulic tests. The In-Pile Test Section(IPS) installed in HANARO FTL is designed as a pressure vessel design conditions of 350℃, 17.5MPa. The instrumentation MI-cables for thermocouples, SPND and LVDT are passed through the sealing plug, which is in the pressure boundary region and is a part of instrumentation feedthrough of MI-cable. In this study, the brazing method and performance test results are introduced to the sealing plug with BNi-2 filler metal, which is selected with consideration of the compatibility for the coolant. The performance was verified through the insulation resistance test, hydrostatic test, and helium leak test.

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 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.