• Nuclear Engineering and Technology
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• v.15 no.4
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• pp.267-279
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• 1983

An analytical system has been established for scrutinizing both uranium- and plutonium-fueled lattices moderated by light water. This system consists of two primary codes. One is a unit cell program called KICC, which has theoretical foundation on the models of GAM and THERMOS incorporated with appropriate approximate treatments for various phenomena, whereas the other is a multi-dimensional diffusion-depletion program entitled KIDD. The adequacy of this system is verified by performing extensive benchmark calculations on a variety of critical experiments. The average value of effective multiplication factors for the selected nineteen UO$_2$ critical experiments of heterogeneous lattice structure is calculated to be 1.0006 with a standard deviation of 0.0039. Power distributions have also been calculated for some critical experiments fueled with both uranium and plutonium of varying concentrations. The maximum percentage difference between the measured and calculated power distributions appears to be less than 5%. This result, together with the previously reported result, illustrates that the KICC/KIDD system is a very effective tool for the analysis of a light water reactor core.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2001.11a
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• pp.19-26
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• 2001

The wear characteristic of Zircaloy-4 tube, which is used for a cladding of light water reactor fuel rod, is investigated experimentally. The experiment is conducted with contacting the crossed tube specimens in air as well as in water at room temperature with various combination of contact normal force and sliding distance of reciprocating motion. The contour and the volume of each wear are examined to study the effect of contact condition and environment on wear. As a result, it is found that the wear volume in the water environment is larger than that in the air for all the contact (i.e., force and sliding distance) conditions. However, the wear depth is greater in air than in water if the contact normal force and the sliding distance are larger. These are explained by the ease of detachment of wear particles from the contact surface. On the other hand, workrate model is applied with the contact shear force range measured by our wear tester. Investigated is the correlation between the workrate and the wear volume increase rate of the present experiment. The parabolic curve is found to fit well for the present wear data.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3709-3715
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• 2023

It is an important work to measure the dimensions of underwater spent fuel assemblies in the nuclear power industry during the overhaul, to judging whether the spent fuel assemblies can continue to be used. In this paper, a three dimensional reconstruction method for underwater spent fuel assemblies of nuclear reactor based on linear structured light is proposed, and the topography and size measurement was carried out based on the reconstructed 3D model. Multiple linear structured light sensors are used to obtain contour size data, and the shape data of the whole spent fuel assembly can be collected by one-dimensional scanning motion. In this paper, we also presented a corrected model to correct the measurement error introduced by lead-glass and water is corrected. Then, we set up an underwater measurement system for spent fuel assembly based on this method. Finally, an underwater measurement experiment is carried out to verify the 3D reconstruction ability and measurement ability of the system, and the measurement error is less than ±0.05 mm.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.276-284
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• 1993

The ABB-CE System-80 reactor power cutback system(RPCS) is designed to enable continuous operation of the reactor without trip in the events of the loss of one of the two main feedwater pumps and loss of load, and thus improves plant availability in a cost effective manner. In this study expansion of RPCS has been investigated for continuous reactor operation without trip in the event of an inward control element assembly(CEA) deviation including a single rod drop. Under the expanded function of RPCS the control system will provide a rapid core power reduction on demand by releasing CEAs to drop into the core and reduce the turbine power, if necessary, to follow the reactor power variation. This design feature which is included as the new design features to be incorporated in the ABB-CE System-80+ meets the EPRI advanced light water reactor(ALWR) requirements. For this study core analysis models of System-80+ have been developed to simulate the nuclear steam supply system(NSSS) response as well as the RPCS initiation of rapid CEA insertion. The results of this study demonstrate that the reactor trip can be avoided in the event of inward CEA deviation including a single rod drop by the RPCS initiation and thus the plant availability and capacity factor would be increased.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.313-342
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• 2014

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.2
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• pp.195-202
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• 2018

As a rule, geological disposal is considered a safe method for final disposal of high-level radioactive waste. However, some long-lived fission products like $^{99}Tc$ and $^{129}I$ contained in spent nuclear fuel are highly mobile as less sorbing anionic species in the subsurface environment and can mainly cause exposure dose to the ecosystem by emission of beta rays in the hundreds of keV range. Therefore, if these two nuclides can be separated and converted with high efficiency into radioactively unharmful nuclides, this would have a positive effect on disposal safety. One candidate method is to transmute these two nuclides in nuclear reactors into short-lived nuclides or into stable nuclides. For this purpose, it is necessary to evaluate which reactor type is more efficient in burning these two nuclides. In this study, the simulation results of nuclear transmutation of $^{99}Tc$ and $^{129}I$ in light water reactor (PWR), heavy water reactor (CANDU) and fast neutron reactor (SFR, MET-1000) are compared and discussed.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.907-914
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• 2015

The energetic steam explosion caused by contact between the high temperature molten core and water is one of the phenomena that may threaten the integrity of the containment vessel during severe accidents of light water reactors (LWRs). We examined the dependence of steam explosion loads in a typical reactor cavity geometry on selected model parameters and initial/boundary conditions by using a steam explosion simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA). Among the parameters, we put an emphasis on the water pool depth that has significance in terms of accident mitigation strategies including cavity flooding. The results showed a strong correlation between the load and the premixed mass, defined as the mass of the molten material in low void zones (void fraction < 0.75). The jet diameter and velocity that comprise the flow rate were the primary factors to determine the premixed mass and the load. The water pool depth also showed a significant impact. The energy conversion ratio based on the enthalpy in the premixed mass was in a narrow range ~4%. Based on this observation, we proposed a simplified method for evaluation of the steam explosion load. The results showed fair agreement with JASMINE.

• 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.52 no.8
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• pp.1603-1610
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• 2020

Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.11-20
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• 2009

Radiation effects on materials are inherently multiscale phenomena in view of the fact that various processes spanning a broad range of time and length scales are involved. A multiscale modeling approach to embrittlement of pressure vessel steels is presented here. The approach includes an investigation of the mechanisms of defect accumulation, microstructure evolution and the corresponding effects on mechanical properties. An understanding of these phenomena is required to predict the behavior of structural materials under irradiation. We used molecular dynamics (MD) simulations at an atomic scale to study the evolution of high-energy displacement cascade reactions. The MD simulations yield quantitative information on primary damage. Using a database of displacement cascades generated by the MD simulations, we can estimate the accumulation of defects over diffusional length and time scales by applying kinetic Monte Carlo simulations. The evolution of the local microstructure under irradiation is responsible for changes in the physical and mechanical properties of materials. Mechanical property changes in irradiated materials are modeled by dislocation dynamics simulations, which simulate a collective motion of dislocations that interact with the defects. In this paper, we present a multi scale modeling methodology that describes reactor pressure vessel embrittlement in a light water reactor environment.