• Structural Engineering and Mechanics
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• 제5권2호
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• pp.149-165
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• 1997

A dynamic time-history analysis of the coupled internals and core in the vertical direction is performed as a part of the fuel assembly qualification program. To reflect the interaction between the fuel rods and grid cage, friction element is developed and is implemented. Also derived here is a method to calculate a hydraulic force on the reactor internals due to pipe break. Peak responses are obtained for the excitations induced from earthquake and pipe break. The dynamic responses such as fuel assembly axial forces and lift-off characteristics are investigated.

• 대한전기학회논문지
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• 제34권12호
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• pp.491-497
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• 1985

An applicaton of the linear optimal control theory to the problem or load follow control in pressurized water reactors (PWR) is investigated. In order to perform the steady-state and load follow operation in PWR, a nonlinear model for the reactor and steam generator is derived and linearized at 50% rated power. Simulation tests are performed for 10% demanded load. Comparing the dynamic response of the newly developed optimal load follow controller with those of the integral error feedback controller proposed by Yang, the rise time of dynamic response of the former is about 15 seconds faster than those of the latter, thus the results indicate that the fast response of the optimal load follow controller is verified. The results of this work are directly applicable to the design of the load follow control systems for commercially operated PWRs.

• 원자력산업
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• 제15권8호통권150호
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• pp.74-79
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• 1995

NPI(Nuclear Power International)사와 그 모기업인 지멘즈(Siemens)사와 프라마톰(Framatome)사는 유럽의 주요 전력업체들의 지원을 받아 차세대원자로인 유럽형 가압경수로(European Pressurized Water Reactor : EPR)를 지금까지 개발해 왔다. EPR은 높은 안전성과 경제성, 평이성, 향상된 운전성 등의 차세대원자로의 당면과제를 만족스럽게 해결함으로써 앞으로 세계 원자력산업에 크게 기여할 것으로 본다. EPR의 개발현황을 살펴본다.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1109-1119
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• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

• Nuclear Engineering and Technology
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• 제15권2호
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• pp.98-109
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• 1983

2차원 다군 확산 이론에 의한 원자로 동특성 방정식의 해를 구하기 위해서 two-step alternating direction explicit method를 도입하였다. Alternating direction implicit method의 특별한 경우로써 이 방법의 정확도 및 안전성을 해석하였다. 이 방법의 타당성을 시험하기 위해서 TWIGL 전산조직에 사용한 implicit difference method와 비교하여 두 방법의 결과가 일치함을 알았다. 이 방법을 이용하여 가압경수형 원자로(PWR)의 제어봉 삽입시의 중성자 신속의 시간변화와, CANDU-PHW 원자로의 가상된 냉각재상실 사고시의 중성자 신속의 시간변화를 계산하여 이들 원자로의 제어능력을 확인하였다.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.327-336
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• 2007

The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields. This paper presents a ROCOM experiment in which water with higher density was injected into a cold leg of the reactor model. Wire-mesh sensors measuring the tracer concentration were installed in the cold leg and upper and lower part of the downcomer. The experiment was run with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water especially for the validation of the Computational Fluid Dynamics (CFD) software ANSYS CFX. A mesh with two million control volumes was used for the calculations. The effects of turbulence on the mean flow were modelled with a Reynolds stress turbulence model. The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

• Nuclear Engineering and Technology
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• 제41권6호
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• pp.765-774
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• 2009

For over 30 years, investigations of the thermohydraulic behavior of pressurized-water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany. The PKL facility models the entire primary side and significant parts of the secondary side of a of pressurized water reactor at a height scale of 1:1. Volumes, power ratings and mass flows are scaled with a ratio of 1:145. The experimental facility consists of four primary loops with circulation pumps and steam generators (SGs) arranged symmetrically around the reactor pressure vessel (RPV). The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermohydraulic phenomena. The PKL tests began in the mid 1970s with the support of the German Research Ministry. Since the mid 1980s, the project has also been significantly supported by the German PWR operators. Since 2001, 25 partner organizations from 15 countries have taken part in the PKL investigations with the support and mediation of the OECD/ NEA (Nuclear Energy Agency). After an overview of PKL history and a short description of the facility, this paper focuses on the investigations carried out since the beginning of the international cooperation, and shows, by means of some examples, what insights can be derived from the tests.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.773-780
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• 2012

The article presented is concerned with an evaluation of the corrosion behavior of SA-508 low alloy steel (LAS) and Type 309L stainless steel (SS) cladding of a reactor pressure vessel under the simulated primary water chemistry of a pressurized water reactor (PWR). The uniform corrosion and galvanic corrosion rates of SA-508 LAS and Type 309L SS were measured in three different control conditions: power operation, shutdown, and power operation followed by shutdown. In all conditions, the dissimilar metal coupling of SA-508 LAS and Type 309L SS exhibited higher corrosion rates than the SA-508 base metal itself due to severe galvanic corrosion near the cladding interface, while the corrosion of Type 309L in the primary water environment was minimal. The galvanic corrosion rate of the SA-508 LAS and Type 309L SS couple measured under the simulated power operation condition was much lower than that measured in the simulated shutdown condition due to the formation of magnetite on the metal surface in a reducing environment. Based on the experimental results, the corrosion rate of SA-508 LAS clad with Type 309L SS was estimated as a function of operating cycle simulated for a typical PWR.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3563-3579
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• 2021

Measured decay heat data of light water reactor (LWR) spent nuclear fuel (SNF) assemblies are adopted to train machine learning (ML) models. The measured data is available for fuel assemblies irradiated in commercial reactors operated in the United States and Sweden. The data comes from calorimetric measurements of discharged pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. 91 and 171 measurements of PWR and BWR assembly decay heat data are used, respectively. Due to the small size of the measurement dataset, we propose: (i) to use the method of multiple runs (ii) to generate and use synthetic data, as large dataset which has similar statistical characteristics as the original dataset. Three ML models are developed based on Gaussian process (GP), support vector machines (SVM) and neural networks (NN), with four inputs including the fuel assembly averaged enrichment, assembly averaged burnup, initial heavy metal mass, and cooling time after discharge. The outcomes of this work are (i) development of ML models which predict LWR fuel assembly decay heat from the four inputs (ii) generation and application of synthetic data which improves the performance of the ML models (iii) uncertainty analysis of the ML models and their predictions.

• 한국소성가공학회:학술대회논문집
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• 한국소성가공학회 2003년도 춘계학술대회논문집
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• pp.327-330
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• 2003

In this paper, a viscoelastic structural analysis for the spent pressurized water reactor(PWR) nuclear fuel disposal canister is carried out to predict the collapse of the canister while the canister is stored in a deep repository for long time. There may exist some subterranean heat in a deep repository while the nuclear fuel disposal canister is stored for long time. Then, a time-dependent viscoelastic structural deformation may occur in the canister due to the subterrnean heat Hence, the viscoelastic stress variation according to time should be computed to predict the structural strength of the canister. A viscoelastic material model is adopted. Analysis results show that even though some subterrnean heat may exist for quite a long time, the canister structure still endures stresses below the yield strength of the canister. Hence, some subterranean heat cannot seriously affect the structural strength of the canister.