• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.90-95
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• 1997

In this work, linear dynamics of a circulating fluid-fueled subcritical reactor system with temperature feedback and external neutron source was modeled and examined. In a circulating fluid-fuel system, the stable region is slightly moved by a circulation fluid effect. The effects of subcriticality and temperature feedback coefficient on the reactor stability were tested by calculating frequency response of neutron density originated from reactivity perturbation or external source oscillation of system. The amplitude transfer function has a different shape near subcritical region due to the exponential term in the transfer function. The results of the study show that at a slightly subcritical region, low frequency oscillation in accelerator current or reactivity can be amplified depending on the temperature feedback. However, as the subcriticality increases, the oscillation becomes negligible regardless of the magnitude of the temperature feedback coefficient.

• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2660-2669
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• 2022

The attractiveness for weapons usage of the proposed fuel cycle for the PT-SCWR was evaluated in this study using the Figure-of-Merit methodology. It was compared to the attractiveness of other fuel cycles namely, Low Enriched Uranium (LEU), U/Th, Re-enriched Reprocessed Uranium (RepU), and Pu/Th/U. The optimal content of natural uranium, which can be added to Pu/Th to render the produced U-233 unattractive, was found to be 9%. A ranking system to compare the attractiveness of the various fuel cycles is proposed. RepU was found to be the most proliferation resistant fuel cycle for the first 100 years,while, the least proliferation resistant fuel cycle was the originally proposed Pu/Th one. The reactivity feedback coefficients were calculated for all proposed fuel cycles. All studied reactivity coefficients have the same sign implying that all the fuel cycles will behave neutronically in a similar way. The Pu/Th/U fuel was found to have the most negative value of the Coolant Void Reactivity which will help to restore the core to a safe status faster in case of a loss-of-coolant accident. The fuel and moderator temperature coefficients did not show significant differences between the fuels studied.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 추계학술발표회요약집
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• pp.167-167
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• 2001

• Nuclear Engineering and Technology
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• 제43권1호
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• pp.63-74
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• 2011

The 600 MWe, pool-type, sodium-cooled, metallic fuel loaded KALIMER-600 (Korea Advanced LiquId MEtal Reactor, 600 MWe) has been conceptually designed with an emphasis on safety by self-regulating (inherent/intrinsic) negative reactivity feedback in the core. Its inherent safety under the ATWS (Anticipated Transient Without Scram) events was demonstrated in an earlier study. Initiating events of an HCDA (Hypothetical Core Disruptive Accident), however, also need to be analyzed for assessment of the margins in the current design. In this study, a hypothetical triple-fault accident, ULOF (Unprotected Loss Of Flow) with a reduced pump halving time, is investigated as an initiator of a core disruptive accident. A ULOF with insufficient primary pump inertia may cause core sodium boiling due to a power-to-flow mismatch. If the positive sodium reactivity resulting from this boiling is not compensated for by other intrinsic negative reactivity feedbacks, the resulting core power burst would challenge the fuel integrity. The present study focuses on determination of the limit of the pump inertia for assuring inherent reactivity feedback and behavior of the core after sodium boiling as well. Transient analyses are performed with the safety analysis code SSC-K, which now incorporates a new sodium boiling model. The results show that a halving time of more than 6.0 s does not allow sodium boiling even with very conservative assumptions. Boiling takes place for a halving time of 1.8 s, and its behavior can be predicted reasonably by the SSC-K.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.544-549
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• 1995

An analysis for Steam Line Break (SLB) events which result in a return-to-power conditions after reactor trip was performed for a postulated Yonggwang Nuclear Power Plant Unit 3 cycle 8. Analysis methodology for post-trip return-to-power SLB is quite different from that of a no return-to-power SLB and is more complicated. Therefore, it is necessary to develop an methodology to analyze the response of the NSSS parameter and the fuel performance for the post-trip return-to-power SLB events. In this analysis, the cases with and without offsite power were simulated by crediting 3-D reactivity feedback effect due to local heatup around stuck CEA and compared with the cases without 3-D reactivity feedback with respect to fuel performance, departure from nucleate boiling ratio (DNBR) and linear heat generation rate (LHGR).

• Nuclear Engineering and Technology
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• 제34권4호
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• pp.286-300
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• 2002

An algorithm for the sodium boiling model has been developed for calculation of the void reactivity feedback as well as the fuel and cladding temperatures in the KALIMER core after onset of sodium boiling. Modeling of sodium boiling in liquid metal reactors using sodium as a coolant is necessary because of phenomenon difference comparing with that observed generally in light water reactor systems. The applied model to the algorithm is the multiple-bubble slug ejection model. It allows a finite number of bubbles in a channel at any time. Voiding is assumed to result from formation of bubbies that (ill the whole cross section of the coolant channel except for the liquid film left on the cladding surface. The vapor pressure, currently, is assumed to be uniform within a bubble The present study is focused on not only demonstration of the vapor bubble behavior predicted by the developed model, but also confirmation of a qualitative acceptance for the model. As a result, the model can represent important phenomena in the sodium boiling, but it is found that further effort is also needed for its completition.

• Nuclear Engineering and Technology
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• 제28권5호
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• pp.452-457
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• 1996

A system, such as a reactor point kinetics equation, can be solved with Adomian Decomposition Method (ADM) which uses the notion that all solutions and operators can be expressed as an infinite sum of those basis states, like Adomian polynomials. In this work, ADM is applied to point reactor kinetics equations for step reactivity insertion, ramp input of reactivity, and nonlinear feedback cases without linearization approximation. The results of ADM are more accurate and faster than those of other existing methods, even though we use comparatively large time step sizes.

• Nuclear Engineering and Technology
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• 제44권2호
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• pp.177-198
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• 2012

This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.

• Nuclear Engineering and Technology
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• 제41권3호
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• pp.307-318
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• 2009

KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2002년도 추계학술발표회요약집
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• pp.157.2-157
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• 2002