• Nuclear Engineering and Technology
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• 제50권5호
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• pp.654-664
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• 2018

Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1871-1880
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• 2020

Round robin analyses for vessel failure probabilities due to PTS events are proposed for plant-specific analyses of all types of reactors developed in Korea. Four organizations, that are responsible for regulation, operation, research and design of the nuclear power plant in Korea, participated in the round robin analysis. The vessel failure probabilities from the probabilistic fracture mechanics analyses are calculated to assure the structural integrity of the reactor pressure vessel during transients that are expected to initiate PTS events. The failure probabilities due to various parameters are compared with each other. All results are obtained based on several assumptions about material properties, flaw distribution data, and transient data such as pressure, temperature, and heat transfer coefficient. The realistic input data can be used to obtain more realistic failure probabilities. The various results presented in this study will be helpful not only for benchmark calculations, result comparisons, and verification of PFM codes developed but also as a contribution to knowledge management for the future generation.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.469-476
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• 2013

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

• Nuclear Engineering and Technology
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• 제43권4호
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• pp.343-354
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• 2011

It is critically important to identify unexpected loose parts in a nuclear reactor pressure vessel, since they may collide with and cause damage to internal structures. Mass estimation can provide key information regarding the kind as well as the location of loose parts. This study proposes a mass estimation method based on an artificial neural network (ANN), which can overcome several unresolved issues involved in other conventional methods. In the ANN model, input parameters are the discrete cosine transform (DCT) coefficients of the auto-power spectrum density (APSD) of the measured impact acceleration signal. The performance of the proposed method is then evaluated through application to a large-sized plate and a 1/8-scaled mockup of a reactor pressure vessel. The results are compared with those obtained using a conventional method, the frequency ratio (FR) method. It is shown that the proposed method is capable of estimating the impact mass with 30% lower relative error than the FR method, thus improving the estimation performance.

• Nuclear Engineering and Technology
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• 제32권4호
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• pp.299-308
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• 2000

A safety assessment of reactor vessel lower head integrity under in-vessel vapor explosion loads has been performed. The core melt relocation parameters were chosen within the ranges of physically realizable bounds. The premixing and explosion calculations were performed using TRACER-II code. Using the calculated explosion pressures imposed on the lower head inner wall, strain calculations were peformed using ANSYS code. Then, the calculated strain results and the established failure criteria were used in determining the failure probability of the lower head, In the explosion analyses, it is shown that the explosion impulses are not altered significantly by the uncertain parameters of triggering location and time, fuel and vapor volume fractions in uniform premixture bounding calculations. Strain analyses show that the vapor explosion-induced lower head failure is not possible under the present framework of assessment. The result of static analysis using the conservative explosion-end pressure of 50 MPa also supports the conclusion. It is recommended, however, that an assessment of fracture mechanics for preexisting cracks be also considered to obtain a more concrete conclusion.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.448-455
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• 2020

The H.B. Robinson Unit 2 (HBR-2) pressure vessel dosimetry benchmark is an in- and ex-Reactor Pressure Vessel (RPV) neutron dosimetry benchmark based on experimental data from the HBR-2 reactor, a 2300-MW PWR designed by Westinghouse and put in operation in March 1971, openly available through the SINBAD Database at OECD/NEA data Bank. The goals of the present work were to carry out three-dimensional (3D) fixed source transport calculations in both Cartesian (X,Y,Z) and cylindrical (R,θ,Z) geometries by using the TORT-3.2 discrete ordinates code on very detailed 3D HBR-2 geometrical models and to test the latest broad-group coupled (47 neutron groups + 20 photon groups) working cross section libraries in FIDO-ANISN format with same structure as BUGLE-96, such as BUGJEFF311.BOLIB, BUGENDF70.BOLIB and BUGLE-B7. The results obtained with all the cited libraries were satisfactory and are here reported and compared.

• 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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• 제38권7호
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• pp.665-674
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• 2006

Three-dimensional neutron flux and fluence of Kori Unit 3 were evaluated using the synthesis technique described in Regulatory Guide 1.190 for all reactor geometry. For this purpose DORT neutron transport calculations from Cycle 1 to Cycle 15 were performed using BUGLE-96 cross-section library. The calculated flux and fluence were validated by comparing the calculated reaction rates to the measurement data from the dosimetry sensor set of the $5^{th}$ surveillance capsule withdrawn at the end of cycle 15 of Kori Unit 3. And then the best estimation of the neutron exposures for the reactor vessel beltline region was performed using the least square evaluation. These results can be used in the assessment of the state of embrittlement of Kori Unit 3 pressure vessel.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.767-790
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• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.290.2-290
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• 2002