• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.104-104
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• 2018

Kori Unit 1 is the first permanent shutdown nuclear power plant in Korea and it is on June 18th, 2017. Spent fuel assemblies began to be discharged from the reactor core to the spent fuel pool(SFP) within one week after shutdown of Kori unit 1 and the campaign was completed on June 27th, 2017. The total number of spent nuclear fuel assemblies in SFP of Kori Unit-1 is 485 and their discharging date is different respectively. So, decay heat was evaluated considering the actual enrichment, operation history and cooling time of the spent fuel assemblies stored in SFP of the Kori Unit-1. The code used in the evaluation is the ORIGEN-based CAREPOOL system developed by KHNP. Decay heat calculation of PWR fuel is based on ANSI/ANS 5.1-2005, "Decay heat power in light water reactors" and ISO-10645, "Nuclear energy - Light water reactors - Calculation of the decay heat power in nuclear fuels. Also, we considered the contribution of fission products, actinide nuclides, neutron capture and radioactive material in decay heat calculation. CAREPOOL system calculates the individual and total decay heat of all of the spent fuel assemblies in SFP of Kori Unit-1. As a result, the total decay heat generated in SFP on June 28th, 2017 when the spent fuel assemblies were discharged from the reactor core, is estimated to be about 4,185.8 kw and to be about 609.5 kw on September 1st, 2018. It was also estimated that 119.6 kw is generated in 2050 when it is 32 years after the permanent shutdown. Figure 1 shows the trend of total decay heat in SFP of Kori Unit-1.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2756-2766
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• 2024

Small rod-controlled pressurized water reactors (PWR) are the ideal energy source for vessel propulsion, benefiting from their high reactivity control efficiency. Since the control rods (CRs) increase the complexity of reactivity control, this paper seeks to study the performance of CRs in small rod-controlled PWRs to extend the lifetime and reduce power offset due to CRs. This study investigates CR grouping, move-in/out strategies, and axially non-uniform design effects on core neutron physics metrics. These metrics include axial offset (AO), core lifetime (CL), fuel utilization (FU), and radial power peaking factor (R-PPF). To simulate the movement of the CRs, a "Critical-CR-burnup" function was developed in OpenMC. In CR designs, the CRs are grouped into three banks to study the simultaneous and prioritized move-in/out strategies. The results show CL extension from 590 effective full power days (EFPDs) to 638-698 EFPDs. A lower-worth prioritized strategy minimizes AO and the extremum values decrease from -0.69 and + 0.81 to -0.28 and + 0.51. Although an axially non-uniform CR design can improve AO at the beginning of cycle (BOC), considering the overall CR worth change is crucial, as a significant decrease can adversely impact axial power distribution during the middle of cycle (MOC).

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.399-411
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• 2003

A model predictive control method is applied to design an automatic controller for thermal power control in a reactor core. The basic concept of the model predictive control is to solve an optimization problem for a finite future at current time and to implement as the current control input only the first optimal control input among the solutions of the finite time steps. At the next time step, the second optimal control input is not implemented and the procedure to solve the optimization problem is then repeated. The objectives of the proposed model predictive controller are to minimize the difference between the output and the desired output and the variation of the control rod position. The nonlinear PWR plant model (a nonlinear point kinetics equation with six delayed neutron groups and the lumped thermal-hydraulic balance equations) is used to verify the proposed controller of reactor power. And a controller design model used for designing the model predictive controller is obtained by applying a parameter estimation algorithm at an initial stage. From results of numerical simulation to check the controllability of the proposed controller at the $5\%/min$ ramp increase or decrease of a desired load and its $10\%$ step increase or decrease which are design requirements, the performances of this controller are proved to be excellent.

• Journal of Mechanical Science and Technology
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• v.14 no.8
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• pp.816-822
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• 2000

In the core of the nuclear power plant of PWR, several cases of fuel failure by unknown causes have been experienced for various fuel types. From the common features of the failure pattern, failure lead time, flow conditions, and flow induced vibration characteristics in nuclear fuel bundles, it is deduced that the fretting wear failure of the fuel rod at the spacer grid position is due to the fluidelastic vibration. In the past, fluidelastic vibration was simulated by quasi -static semi-analytical model, so called the static model, which could not account for the interaction between the rods within a bundle. To overcome this defect and to provide for more flexibilities applicable to the fuel bundle, Tanaka's unsteady model was modified to accomodate the geometrical differences and governing parameter changes during the operations such as the number of rods, pitch to diameter ratio (P/D), spring force, damping coefficient, etc. The critical velocity was calculated by solving the governing equations with the MATLAB code. A comparison between the estimated critical velocity and the test result shows a good agreement. Finally, the level of decrease of the critical velocity due to the reduction in the spring force and reduced damping coefficient due to the radiation exposure is also estimated.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.383-387
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• 2004

The Fuel Test Loop (FTL) consists of In-Pile Test Section (IPS) and Out-of-Pile System (OPS). The test condition in IPS such as pressure, temperature and quality of the main cooling water, can be controlled by the OPS. The FTL has been developed to be able to irradiate three pins to the core irradiation hole (IR1 hole) by considering for its utility and user's irradiation requirement. The IPS vessel assembly (IVA) consists of IPS head, outer pressure vessel, inner pressure vessel, inner assembly and test fuel carrier. The IVA is approximately 5.6 m long and fits within a 74 mm in diameter envelope over the full height of the chimney. Above the top of the chimney, the head of the IPS is enlarged to allow the closure flanges and pipe work connections. IVA was designed to test the CANDU and PWR nuclear fuel pin together. Specially, wished to minimize interference by nuclear fuel change in design and synthesize these items and shape design for IVA.

• Nuclear Engineering and Technology
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• v.22 no.2
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• pp.95-100
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• 1990

The exact expression for the 1151 force on a fuel assembly in a reactor core is derived in terms of calculable hydraulic parameters. The relation for the lift force. pressure drop, buoyancy force, viscous force. and fuel assembly weight is discussed. Based on the derived exact expression. error analysis is made for a simple expression applying COBRA IV-i to a typical PWR fuel assembly. The error analysis revealed that the error of the simple expression consists of four terms and the overall error depends on the flow rate change direction, and its magnitude is about 1%.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1854-1858
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• 2007

The reliability assurance with respect to the test procedure and results of the out-pile mechanical performance test for the nuclear fuel assembly is an essential task to assure the test quality and to get a permission for fuel loading into the commercial reactor core. For the case of vibration test, which is carried out to obtain basic dynamic characteristics of the fuel assembly, proper management and appropriate calibration of instruments and devices used in the test, various efforts to minimize the possible error during the test and signal acquisition process are needed. Additionally, the deep understanding both of the theoretical assumption and simplification cation for the signal processing/modal analysis and of the functions of the devices used in the test were highly required. Finally, to verify the test result to represent the accurate natural characteristics of the structure, the proper correlation analysis between the theoretical and experimental method has to be carried out. In this study, the overall procedure and result of lateral vibration test for the fuel assembly's mechanical characterization were briefly introduced. A series of measures to assure and improve the reliability of the vibration test were discussed.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.989-994
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• 2009

We at Mitsubishi have lined up Gen-III+ solutions for a wide variety of global customers: ATMEA1 of the 1100MWe class, and an APWR with the largest capacity of 1700MWe. In this paper, we would like to introduce the APWR. With an increased requirement for nuclear power generation as an effective countermeasure against global warming, we have established the APWR plant, a large-capacity Mitsubishi standard reactor combining our accumulated experience and technology as an integrated PWR plant supplier. The APWR plant has achieved high reliability, safety and enhanced economy based on a technology that has been developed with the support of the government and utilities through improvement and standardization programs of light water reactors. Currently, Tsuruga Units 3 and 4, the first two APWRs, are undergoing licensing, while we are making efforts to obtain the standard design certification (DC) of US-APWR and preparing for the European Utility Requirements (EUR) compliance assessment of EU-APWR. Mitsubishi Heavy Industries, Ltd. (MHI) positions the APWR as a core technology that will contribute to the prevention of global warming and meet worldwide requirements.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1096-1108
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• 2016

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U-transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.243-256
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• 2011

This paper considers the concept of analog computing based on a cellular neural network (CNN) paradigm to simulate nuclear reactor dynamics using a time-dependent second order form of the neutron transport equation. Instead of solving nuclear reactor dynamic equations numerically, which is time-consuming and suffers from such weaknesses as vulnerability to transient phenomena, accumulation of round-off errors and floating-point overflows, use is made of a new method based on a cellular neural network. The state-of-the-art shows the CNN as being an alternative solution to the conventional numerical computation method. Indeed CNN is an analog computing paradigm that performs ultra-fast calculations and provides accurate results. In this study use is made of the CNN model to simulate the space-time response of scalar flux distribution in steady state and transient conditions. The CNN model also is used to simulate step perturbation in the core. The accuracy and capability of the CNN model are examined in 2D Cartesian geometry for two fixed source problems, a mini-BWR assembly, and a TWIGL Seed/Blanket problem. We also use the CNN model concurrently for a typical small PWR assembly to simulate the effect of temperature feedback, poisons, and control rods on the scalar flux distribution.