• Nuclear Engineering and Technology
• /
• 제49권4호
• /
• pp.752-759
• /
• 2017

The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

• Nuclear Engineering and Technology
• /
• 제49권1호
• /
• pp.71-81
• /
• 2017

When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

• 한국소음진동공학회:학술대회논문집
• /
• 한국소음진동공학회 2005년도 추계학술대회논문집
• /
• pp.267-270
• /
• 2005

Finite element model of pressurized water reactor internals were obtained using ANSYS software package to analyze dynamic characteristics. The pressure vessel, hold-down ring, alinement key, core support barrel(CSB), upper guide structure(UGS) and fluid gap were fully modeled using structural solid element(SOLID45) and fluid element(FLUID80) which is one of element types. Also modal analysis using the above finite element model has been performed. As a result, it was found that the fundamental beam mode natural frequency of the CSB were 8.2 Hz, the shell mode one 14.5 Hz. To verify the Finite Element Analysis(FEA), we compare the analysis result with experimental data that is obtained from the plant IVMS(internal Vibration Monitoring System). The experimental results are good agreement with the FEA model.

• Nuclear Engineering and Technology
• /
• 제47권7호
• /
• pp.924-933
• /
• 2015

The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional $^{235}U$ burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using $^{233}U$, $^{242}Pu$, $^{150}Nd$, and $^{133}Cs$ as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

• 기계저널
• /
• 제51권12호
• /
• pp.51-54
• /
• 2011

이 글에서는 제4세대 원자로로 다시 부각되고 있는 소듐냉각고속로(SFR: Sodium-cooled Fast Reactor)의 활용성, 계통설계 구성 및 공학적 안전설비에 대하여 가압경수로(PWR: Pressurized Water Reactor)와의 차이점을 위주로 소개한다.

• 소음진동
• /
• 제7권5호
• /
• pp.837-845
• /
• 1997

본 연구에서는 가압경수로의 핵심부품인 원자로용기와 제어봉구동장치사이의 동적 상호작용의 영향을 평가하였다. 원자로용기와 제어봉구동장치를 단순 수학모델화하여 단순조화 기저가진에 대한 정상상태 주파수응답을 구하고, 응답을 최소화할 수 있는 설계변수를 제시하였다. 단순 수학모델의 적합성을 입증하기 위하여 원자로용기와 제어봉구동장치의 유한요소 모델에 대하여 ANSYS 코드를 사용하여 해석한 후 정상상태응답을 수학모델로 구한 응답과 비교하였다. 또한 기존설계와 최적화설계에 대하여 지진사고 시간이력해석을 각각 수행함으로써, 본 연구에서 제시한 최적설계변수가 내진설계에 대하여도 적용할 수 있음을 확인하였다.

• Nuclear Engineering and Technology
• /
• 제38권5호
• /
• pp.411-416
• /
• 2006

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the comer 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

• Nuclear Engineering and Technology
• /
• 제45권5호
• /
• pp.589-596
• /
• 2013

In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ~2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

• Nuclear Engineering and Technology
• /
• 제29권5호
• /
• pp.375-384
• /
• 1997

Relative power density distributions of the Kori Unit 1 pressurized water reactor are calculated by Monte Carlo modeling with the MCNP code. The Kori Unit 1 core is modeled on a three-dimensional representation of the one-eighth of the reactor in-vessel component with reflective boundaries at 0 and 45 degrees. The axial core model is based on half core symmetry and is divided into four axial segments. Fission reaction density in each rod is calculated by following 100 cycles with 5,000 test neutrons in each cycle after starling with a localized neutron source and ten noncontributing settle cycles. Relative assembly power distributions are calculated from fission reaction densities of rods in assembly. After 100 cycle calculations, the system converges to a k value of 1.00039 $\geq$ 0.00084. Relative assembly power distribution is nearly the same with that of the Kori Unit 1 FSAR. Applicability of the full-scope Monte Carlo simulation in the power distribution calculation is examined by the relative root moan square error of 2.159%.

• Nuclear Engineering and Technology
• /
• 제3권2호
• /
• pp.59-64
• /
• 1971

각 상태변수의 출력에서 일정한 이득의 신호를 다시 궤환시키는 최선의 제어방법을 실제에 적용시켜 보았다. 이러한 궤환 신호의 값은 계 전체의 전달함수를 구하고 우리가 원하는 계의 전달함수와 비교하여 s의 같은 차수끼리 서로 같도록 놓음으로서 결정하였다. 이 방법은 복잡한 원자로의 궤환 계수를 매우 간단하게 결정할 수 있으며 특히 PWR의 제어에 간편 방법으로 사용하여 보았다.