• Nuclear Engineering and Technology
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• 제50권5호
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• pp.648-653
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• 2018

Elimination of soluble boron from reactor design eliminates boron-induced reactivity accidents and leads to a more negative moderator temperature coefficient. However, a large negative moderator temperature coefficient can lead to large reactivity feedback that could allow the reactor to return to power when it cools down from hot full power to cold zero power. In soluble boron-free small modular reactor (SMR) design, only control rods are available to control such rapid core transient. The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model. The control rod worth requirement obtained from the reactivity balance is a minimum control rod worth to maintain the reactor critical during the whole cycle. The minimum available rod worth must be larger than the control rod worth requirement to manipulate the reactor safely in each reactor state. It is found that the SMR does have enough control rod worth available during rapid transient to maintain the SMR at subcritical below k-effectives of 0.99 for both hot zero power and cold zero power.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.77-82
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• 1997

The CANDU-6 transition core refuelled from 37-element fuel to CANFLEX-NU fuel has been evaluated by an 100full power day time-dependent fuel-management simulation to find the core compatibility with the CANFLEX fuel loading. The simulation calculations for the transition core were carried out with the RFSP code, provided by the cell averaged fuel properties obtained from the POWDERPUFS-V code. The simulation results were compared with those of the current 37-element fuel loading only. The results show that the CANFLEX-NU fuel bundles will be compatible with the CANDU-6 reactor because the core physics characteristics of CANFLEX-NU fuel are very similar to those of the 37-element fuel bundle.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.152-161
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• 2022

We analyzed mitigation strategies for steam generator tube rupture (SGTR) accidents using MARS code under both full-power and low-power and shutdown (LPSD) conditions. In general, there are two approaches to mitigating SGTR accidents: supplementing the reactor coolant inventory using safety injection systems and depressurizing the reactor coolant system (RCS) by cooling it down using the intact steam generator. These mitigation strategies were compared from the viewpoint of break flow from the ruptured steam generator tube, the core integrity, and the possibility of the main steam safety valves opening, which is associated with the potential release of radiation. The "cooldown strategy" is recommended for break flow control, whereas the "RCS make-up strategy" is better for RCS inventory control. Under full power, neither mitigation strategy made a significant difference except for on the break flow while, in LPSD modes, the RCS cooldown strategy resulted in lower break and discharge flows, and thus less radiation release. As a result, using the cooldown strategy for an SGTR under LPSD conditions is recommended. These results can be used as a fundamental guide for mitigation strategies for SGTR accidents according to the operational mode.

• Nuclear Engineering and Technology
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• 제43권6호
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• pp.583-592
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• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• 시스템엔지니어링학술지
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• 제19권2호
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• pp.46-58
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• 2023

Deterministic safety analysis is a crucial part of safety assessment, particularly when it comes to demonstrating the safety of nuclear power plant designs. The traditional approach to deterministic safety analysis models is to model the nuclear core using point kinetics. However, this simplified approach does not fully reflect the real core behavior with proper moderator and fuel reactivity feedbacks during the transient. The use of Multi-Physics approach allows more precise simulation reflecting the inherent three-dimensionality (3D) of the problem by representing the detailed 3D core, with instantaneous updates of feedback mechanisms due to changes of important reactivity parameters like fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC). This paper addresses a CEA ejection accident at hot full power (HFP), in which the underlying strong and un-symmetric feedback between thermal-hydraulics and reactor kinetics exist. For this purpose, a multi-physics analysis tool has been selected with the nodal kinetics code, 3DKIN, implicitly coupled to the thermal-hydraulic code, RELAP5, for real-time communication and data exchange. This coupled approach enables high fidelity three-dimensional simulation and is therefore especially relevant to reactivity initiated accident (RIA) scenarios and power distribution anomalies with strong feedback mechanisms and/or un-symmetrical characteristics as in the CEA ejection accident. The Systems Engineering approach is employed to provide guidance in developing the work in a systematic and efficient fashion.

• Nuclear Engineering and Technology
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• 제52권2호
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• pp.230-237
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• 2020

In this part, an implicit time dependent solution is presented for the Boltzmann transport equation discretized by the analytic coarse mesh finite difference method (ACMFD) over the spatial domain as well as the simplified P₃ (SP₃) for the angular variable. In the first part of this work we proposed a SP₃-ACMFD approach to solve the static eigenvalue equations which provide the initial conditions for temp dependent equations. Having solved the 3D multi-group SP₃-ACMFD static equations, an implicit approach is resorted to ensure stability of time steps. An exponential behavior is assumed in transverse integrated equations to establish a relationship between flux moments and currents. Also, analytic integration is benefited for the time-dependent solution of precursor concentration equations. Finally, a multi-channel one-phase thermal hydraulic model is coupled to the proposed methodology. Transient equations are then solved at each step using the GMRES technique. To show the sufficiency of proposed transient SP₃-ACMFD approximation for a full core analysis, a comparison is made using transport peers as the reference. To further demonstrate superiority, results are compared with a 3D multi-group transient diffusion solver developed as a byproduct of this work. Outcomes confirm that the idea can be considered as an economic interim approach which is superior to the diffusion approximation, and comparable with transport in results.

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3059-3072
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• 2022

A Jacobian-Free Newton Krylov Two-Nodal Coarse Mesh Finite Difference algorithm based on Nodal Expansion Method (NEM_TNCMFD_JFNK) is successfully developed and proposed to solve the three-dimensional (3D) and multi-group reactor physics models. In the NEM_TNCMFD_JFNK method, the efficient JFNK method with the Modified Incomplete LU (MILU) preconditioner is integrated and applied into the discrete systems of the NEM-based two-node CMFD method by constructing the residual functions of only the nodal average fluxes and the eigenvalue. All the nonlinear corrective nodal coupling coefficients are updated on the basis of two-nodal NEM formulation including the discontinuity factor in every few newton steps. All the expansion coefficients and interface currents of the two-node NEM need not be chosen as the solution variables to evaluate the residual functions of the NEM_TNCMFD_JFNK method, therefore, the NEM_TNCMFD_JFNK method can greatly reduce the number of solution variables and the computational cost compared with the JFNK based on the conventional NEM. Finally the NEM_TNCMFD_JFNK code is developed and then analyzed by simulating the representative PWR MOX/UO₂ core benchmark, the popular NEACRP 3D core benchmark and the complicated full-core pin-by-pin homogenous core model. Numerical solutions show that the proposed NEM_TNCMFD_JFNK method with the MILU preconditioner has the good numerical accuracy and can obtain higher computational efficiency than the NEM-based two-node CMFD algorithm with the power method in the outer iteration and the Krylov method using the MILU preconditioner in the inner iteration, which indicates the NEM_TNCMFD_JFNK method can serve as a potential and efficient numerical tool for reactor neutron diffusion analysis module in the JFNK-based multiphysics coupling application.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.581-588
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• 2013

This paper deals with the Safety Analysis for $CANDU^{(R)}$ 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2014년도 추계학술대회 논문집
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• pp.106-109
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• 2014

In order to obtain dynamic behaviors of complex structures, it demands large amounts computational cost and time to perform the numerical analysis. The model reduction method helps these problems by dividing the full model into primary and unnecessary parts. In this research, we perform the modal analysis using the dynamic substructuring method, which is one of the model reduction methods, in order to obtain the dynamic characteristics of the cylindrical structures efficiently. To select the master degrees of freedom (dofs), we consider the mode shapes of the cylindrical structures. And then, we identify the validity of the dynamic substructuring method by applying the method to the simple cylinder and core support barrel (CSB) which is one of the reactor internals with the cylindrical shape. The results demonstrate that the dynamic characteristics from the dynamic substructuring method are well matched with the original method.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.185-190
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• 1996

RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the ＊SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.