• 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.42 no.4
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• pp.434-441
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• 2010

The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant.

• Nuclear Engineering and Technology
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• v.40 no.4
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• pp.269-276
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• 2008

Operational license of WWER-440/213 units at Paks NPP, Hungary is limited to the design lifetime of 30 years. Prolongation by additional 20 years of the operational lifetime is feasible. Moreover, enhancement of the reactor thermal power by 8% will increase both the net power output and the competitiveness of the plant. Paks NPP is a pioneer considering the power up-rate and preparation of long-term operation of WWER-440/213 design. Systematic preparatory work for long-term operation of Paks NPP has been started in 2000. A regulatory framework and a comprehensive engineering practice have been developed. According to the authors view, creation of a gapless engineering system via consequent application of best practices, and feed-back of experiences together with proper consideration of WWER-440/V213 features are the decisive elements of ensuring the safety of long-term operation. That systematic engineering approach is in the focus of recent paper. Key elements of justification and measures for ensuring the safety of long-term operation of Paks NPP WWER-440/213 units are identified and discussed. These are the assessment of plant condition and review of adequacy of ageing management programmes, also the review, validation and reconstitution of time limited ageing analyses as core tasks of licence renewal.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1479-1486
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• 2019

The purpose of the present study is to develop the 3D static and noise simulator based on Galerkin Finite Element Method (GFEM) using the unstructured hexahedral elements. The 3D, 2G neutron diffusion and noise equations are discretized using the unstructured hexahedral by considering the linear approximation of the shape function in each element. The validation of the static calculation is performed via comparison between calculated results and reported data for the VVER-1000 benchmark problem. A sensitivity analysis of the calculation to the element type (unstructured hexahedral or tetrahedron elements) is done. Finally, the neutron noise calculation is performed for the neutron noise source of type of variable strength using the Green function technique. It is shown that the error reduction in the static calculation is considerable when the unstructured tetrahedron elements are replaced with the hexahedral ones. Since the neutron flux distribution and neutron multiplication factor are appeared in the neutron noise equation, the more accurate calculation of these parameters leads to obtaining the neutron noise distribution with high accuracy. The investigation of the changes of the neutron noise distribution in axial direction of the reactor core shows that the 3D neutron noise analysis is required instead of 2D.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.13-30
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• 2019

This paper presents the recent improvements in the DeCART code for HTGR analysis. A new 190-group DeCART cross-section library based on ENDF/B-VII.0 was generated using the KAERI library processing system for HTGR. Two methods for the eigen-mode adjoint flux calculation were implemented. An azimuthal angle discretization method based on the Gaussian quadrature was implemented to reduce the error from the azimuthal angle discretization. A two-level parallelization using MPI and OpenMP was adopted for massive parallel computations. A quadratic depletion solver was implemented to reduce the error involved in the Gd depletion. A module to generate equivalent group constants was implemented for the nodal codes. The capabilities of the DeCART code were improved for geometry handling including an approximate treatment of a cylindrical outer boundary, an explicit border model, the R-G-B checker-board model, and a super-cell model for a hexagonal geometry. The newly improved and implemented functionalities were verified against various numerical benchmarks such as OECD/MHTGR-350 benchmark phase III problems, two-dimensional high temperature gas cooled reactor benchmark problems derived from the MHTGR-350 reference design, and numerical benchmark problems based on the compact nuclear power source experiment by comparing the DeCART solutions with the Monte-Carlo reference solutions obtained using the McCARD code.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1079-1087
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• 2021

NTP-ERSN is a package developed for solving the multigroup form of the discrete ordinates, characteristics and collision probability of the Boltzmann transport equation in one-dimensional cartesian geometry, by combining pin cells. In this work, C5G7 MOX benchmark is used to verify the accuracy and efficiency of NTP-ERSN package, by treating reactor core problems without spatial homogenization. This benchmark requires solutions in the form of normalized pin powers as well as the vectors and the eigenvalue. All NTP-ERSN simulations are carried out with appropriate spatial and angular approximations. A good agreement between NTP-ERSN results with those obtained with OpenMC calculation code for seven energy groups. In addition, our studies about angular and mesh refinements are carried out to produce better quality solution. Moreover, NTP-ERSN GUI has also been updated and adapted to python 3 programming language.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.793-803
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• 2021

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

• Journal of the Korean Society of Systems Engineering
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• v.16 no.2
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• pp.1-15
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• 2020

Nuclear power plants in South Korea are operated to cover the baseload demand. Hence they are operated at 100% rated power and do not deploy power tracking control except for startup, shutdown, or during transients. However, as the contribution of renewable energy in the energy mix increases, load follow operation may be needed to cover the imbalance between consumption and production due to the intermittent nature of electricity produced from the conversion of wind or solar energy. Load follow operation may be quite challenging since the operators need to control the axial power distribution and core reactivity while simultaneously conducting the power maneuvering. In this paper, a systems engineering approach for multi-physics load follow simulation of APR1400 is performed. RELAP5/SCDAPSIM/MOD3.4/3DKIN multi-physics package is selected to simulate the Korean Advanced Power Reactor, APR1400, under load follow operation to reflect the impact of feedback signals on the system safety parameters. Furthermore, the systems engineering approach is adopted to identify the requirements, functions, and physical architecture to provide a set of verification and validation activities that guide this project development by linking each requirement to a validation or verification test with predefined success criteria.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2650-2659
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• 2022

The dense granular flow spallation target is a new target concept proposed for an Accelerator-Driven Sub-critical (ADS) system. In this paper, the beam-target configurations of a tungsten granular flow target for the ADS with a thermal power of 1 GW is explored. The beam profile options using different scanning methods are discussed. The critical geometry parameters are adjusted to investigate the performance of the granular target from the aspects of neutron efficiency, stability and temperature distribution in target medium. To figure out how the target under accident conditions would behave, different clogging conditions are induced in the simulation. The dynamic processes are analyzed and some important parameters such as abnormal temperature rise and beam cutoff time window are obtained. The response of the sub-critical reactor to a clogging accident is also investigated. It is indicated that the monitoring of the granular flow by the neutron detectors in the sub-critical core will be effective.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.138-151
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• 2022

Cladding ballooning and rupture are the important phenomena at the early stage of a severe accident. Most severe accident analysis codes determine the cladding rupture based on simple parameter models. In this paper, a FRTMB module was developed using the thermal-mechanical model to analyze the fuel mechanical behavior. The purpose is to judge the cladding rupture with the severe accident analysis code. The FRTMB module was integrated into the self-developed severe accident analysis code ISAA to simulate the PHEBUS FPT3 experiment. The predicted rupture time and temperature of the cladding were basically consistent with the measured values, which verified the correctness and effectiveness of the FRTMB module. The results showed that the rising of gas pressure in the fuel rod and high temperature led to cladding ballooning. Consequently, the cladding hoop strain exceeded the strain limit, and the cladding burst. The developed FRTMB module can be applied not only to rod-type fuel, but also to plate-type fuel and other types of reactor fuel rods. Moreover, the FRTMB module can improve the channel blockage model of ISAA code and make contributions to analyzing the effect of clad ballooning on transient and subsequent parts of core degradation.