• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.29 no.10
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• pp.538-550
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• 2017

In a PWR (Pressurized Water Reactor), the appropriate heat removal from the surface of fuel rod bundle is important for ensuring thermal margins and safety. Although many CFD (Computational Fluid Dynamics) software have been used to predict complex flows inside fuel assemblies with mixing vanes, there is no domestic regulatory guideline for the comprehensive evaluation of CFD software. Therefore, from the nuclear regulatory perspective, it is necessary to perform the systematic assessment and prepare the domestic regulatory guideline for checking whether valid CFD software is used for nuclear safety problems. In this study, to provide systematic evaluation and guidance on the applicability of CFD software to the domestic nuclear safety area, the results of the sensitivity analysis for the effect of the discretization scheme accuracy for the convection terms and turbulence models, which are main factors that contribute to the uncertainty in the calculation of the nuclear safety problems, on the prediction performance for the turbulent flow distribution inside the fuel assembly with split-type mixing vanes were explained.

• The KSFM Journal of Fluid Machinery
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• v.5 no.3 s.16
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• pp.25-32
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• 2002

Structural and dynamic analyses of inducer and impeller for an oxidizer turbopump are peformed to investigate the safety level of strength and vibration at a design point. Due to high rotational speed of turbopump, effects of centrifugal forces are carefully considered in the structural analysis. Hydrodynamic pressure is also considered as an external force applied to inducer and impeller blades. A three-dimensional Finite Element Method (FEM) is used for linear and nonlinear structural analyses with modified Newton-Raphson iteration method. After the nonlinear trim solution is obtained from the structural analysis, dynamic characteristics are obtained as a function of rotational speed from the linearized eigenvalue analysis at an equilibrium position. According to the results of numerical analysis, the safety margins of strength and vibration resonances are sufficient enough for safe operation within the requited life cycle.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.592-597
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• 1996

The KALIMER core, of which nuclear design is largely governed by inherent safety and reactivity control issues, is fueled with metallic fuel, and the initial core will be loaded with 20% enriched Uranium metal fuel. KALIMER safety design objectives include the accommodation of unprotected, ATWS events without operator action, and without the support of active shutdown, shutdown heat removal, or any automatic system without damage to the plant and without jeopardizing public safety. The transient analysis of the core designs has been focused on severe events to assess the margins in the design, and ATWS events are the most severe events that must be accommodated by the KALIMER design. The ATWS performance has been evaluated for the preliminary initial core design of KALIMER with a particular emphasis on the inherent negative reactivity feedback effects, including the Doppler, sodium density, fuel axial expansion, core radial expansion, and control rod driveline expansion. Results show that the Uranium metal core design meets the temperature limits with margin.

• 유체기계공업학회:학술대회논문집
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• 2001.11a
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• pp.271-278
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• 2001

Structural and dynamic analyses of inducer and impeller for a oxidizer turbopump are peformed to investigate the safety level of strength and vibration at design point. Due to high rotational speed of turbopump, effects of centrifugal forces are carefully considered in the structural analysis. Hydrodynamic pressure is also considered as an external force applied to inducer and impeller blades. A three dimensional finite element method(FEM) is used for linear and nonlinear structural analyses with modified Newton-Raphson iteration method. After the nonlinear trim solution is obtained from the structural analysis, dynamic characteristics are obtained as a function of rotational speed from the linearized eigenvalue analysis at an equilibrium position. According to the results of numerical analysis, the safety margins of strength and vibration resonances m sufficient enough to be operated safely within the required life cycle.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.54-59
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• 2019

The main objective of nuclear reactor safety is to maintain the nuclear fuel in a thermally safe condition with enough safety margins during normal operation and anticipated operational occurrences. In this research, core flow bypass is studied under the conditions of the unavailability of safety systems. As core bypass occurs, the core flow rate is assumed to decrease exponentially with a time constant of 25 s to new steady state values of 20, 40, 60, and 80% of the nominal core flow rate. The thermal hydraulic code PARET is used through these calculations. Reactor thermal hydraulic stability is reported for all cases of core flow bypass.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.489-494
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• 1997

A comparative transient analyses were performed for oxide and nitride cores or a large (3000 MWt), pool-type, liquid-metal-cooled reactor (LMR). The study was focused on three representative accident initiators with failure to scram : the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected fast transient overpower (UFTOP). The margins to fuel melting and sodium boiling have been evaluated for these representative transients. The results show that there is an increase in safety margin with nitride core which maintains the physical dimensions of the oxide core.

• International Journal of Aerospace System Engineering
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• v.9 no.1
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• pp.1-15
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• 2022

Pyrotechnic initiators provide a source of pyrotechnic energy used to initiate a variety of space mechanisms. Pyrotechnic systems build in electromagnetic environment that may lead to critical or catastrophic hazards. Special precautions are need to prevent a pulse large enough to trigger the initiator from appearing in the pyrotechnic firing circuits at any but the desired time. The EMC verification shall be shown by analysis or test that the pyrotechnic systems meets the requirements of inadvertent activation. The MIL-STD-1576 and two range safeties, AFSPC and CSG, require the safety margin for electromagnetic potential hazards to pyrotechnic systems to a level at least 20 dB below the maximum no-fire power of the EED. The PC23 is equivalent to NASA standard initiator and the 1EPWH100 squib is ESA standard initiator. This paper verifies the two safety margins for electromagnetic potential hazards. The first is verified by analyzing against a RF power. The second is verified by testing against a DC current. The EMC safety margin requirement against RF power has been demonstrated through the electric field coupling analysis in differential mode with 21 dB both PC23 and 1EPWH100, and in common mode with 58 dB for PC23 and 48 dB for 1EPWH100 against the maximum no-fire power of the EED. Also, the EMC safety margin requirement against DC current has been demonstrated through the electrical isolation test for the pyrotechnic firing circuits with greater than 20 dB below the maximum no-fire current of the EED.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.3
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• pp.681-689
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• 1994

A model to predict the onset of significant void (OSV) in vertical flow between parallel plates has been developed. The model was compared to the experimental data of Whittle and Forgan (1967) and Dougherty et al. (1990), showing excellent agreement. The model was also compared with the Saha-Zuber(1974) correlation, which has been widely used in computer codes for nuclear safety analysis. The present theory is more conservative than this correlation, and further shows that, contrary to this correlation, the Stanton number is not solely related to the Peclet number. This may explain the large error margins required for the Saha-Zuber correlation, and also the scatter beyond the error margins specified by the authors. The steady-state OSV heat fluxes for equal and unequal heating cases between parallel plates were compared. The arithmetic mean of heat fluxes for unequal heating cases is less than the heat flux for equal heating cases. The result may imply that OSV is controlled by local thermal parameters rather than bulk parameters.

• The Transactions of the Korean Institute of Electrical Engineers A
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• v.53 no.3
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• pp.129-134
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• 2004

The Available Transfer Capability (ATC) is defined as the measure of the transfer capability remaining in the physical transmission network for further commercial activity above already committed uses. The ATC determination s related with Total Transfer Capability (TTC) and two reliability margins-Transmission Reliability Capability (TRM) and Capacity Benefit Margin(CBM) The TRM is the component of ATC that accounts for uncertainties and safety margins. Also the TRM is the amount of transmission capability necessary to ensure that the interconnected network is secure under a reasonable range of uncertainties in system conditions. The CBM is the translation of generator capacity reserve margin determined by the Load Serving Entities. This paper describes a method for determining the TTC and TRM to calculate the ATC in the Bulk power system (HL II). TTC and TRM are calculated using Power Transfer Distribution Factor (PTDF). PTDF is implemented to find generation quantifies without violating system security and to identify the most limiting facilities in determining the network’s TTC. Reactive power is also considered to more accurate TTC calculation. TRM is calculated by alternative cases. CBM is calculated by LOLE. This paper compares ATC and TRM using suggested PTDF with using CPF. The method is illustrated using the IEEE 24 bus RTS (MRTS) in case study.