• Nuclear Engineering and Technology
• /
• 제56권3호
• /
• pp.873-879
• /
• 2024

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.

• 공업화학
• /
• 제9권2호
• /
• pp.268-272
• /
• 1998

액체금속로 냉각재인 액체 소듐에서 시편의 누출특성을 소듐-물 반응 실험에 의해 조사하였다. 소듐-물 반응 현상의 확인은 물 누출 실험 전후에 Fe, Cr 및 Ni 등과 같은 시편의 조성 변화로 확인하였다. $100kg/cm^2$의 누출 압력으로 4시간 동안 시편의 누출 경로를 통해 물을 누출시킨 결과, 누출경로에서 소듐-물 반응생성물들이 침적되어 있는 것을 확인하였으나, 부식에 의해 누출경로가 완전 파열되어 다량의 수증기가 액체 소듐속으로 빠져나가는 re-openning 현상은 관찰되지 않았다. 시편의 누출경로가 막히는 self-plugging 현상은 소듐-물 반응에 의한 반응생성물과 시편의 부식에 의한 부식 생성물이 주 원인으로 추정되고, re-openning 현상은 시편의 누출경로에서 열적인 transient로 추정되었다.

• Nuclear Engineering and Technology
• /
• 제41권6호
• /
• pp.785-796
• /
• 2009

A KALIMER-600 concept which is a type of sodium-cooled fast reactor, has been developed at KAERI. It uses sodium as a primary coolant and is a pool-type reactor to enhance safety. Also, a supercritical carbon dioxide ($CO_2$) Brayton cycle is considered as an alternative to an energy conversion system to eliminate the sodium water reaction and to improve efficiency. In this study, a simplified model for analyzing the thermodynamic performance of the KALIMER-600 coupled with a supercritical $CO_2$ Brayton cycle was developed. To develop the analysis model, a commercial modular modeling system (MMS) was adopted as a base engine, which was developed by nHance Technology in USA. It has a convenient graphical user interface and many component modules to model the plant. A new user library for thermodynamic properties of sodium and supercritical $CO_2$ was developed and attached to the MMS. In addition, some component modules in the MMS were modified to be appropriate for analysis of the KALIMER-600 coupled with the supercritical $CO_2$ cycle. Then, a simplified performance analysis code was developed by modeling the KALIMER-600 plant with the modified MMS. After evaluating the developed code with each component data and a steady state of the plant, a simple power reduction and recovery event was evaluated. The results showed an achievable capability for a performance analysis code. The developed code will be used to develop the operational strategy and some control logics for the operation of the KALIMER-600 with a supercritical $CO_2$ Brayton cycle after further studies of analyzing various operational events.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 1999년도 추계학술대회 논문집 전문대학교육위원 P
• /
• pp.1-3
• /
• 1999

EM (ElectroMagnetic) pump is used for the purpose of transporting liquid sodium coolant with electrical conductivity in the LMR(Liquid Metal Reactor). (In the present study, pilot EM pump has been designed by using of equivalent circuit method which is commonly employed to analyze linear induction machines for the test of removal of residual heat. The length and diameter of the pump have fixed values of 840 mm and 101.6 mm each by taking account of geometrical size of circulation loop for the installation of EM pump. Flowrate versus developing pressure is related from Laithwaite's standard design formula and the characteristic analyses of developing force and efficiency are carried out according to change of input frequency. From the characteristic curve, input frequency of 13 Hz is determined as the design frequency. On the other hand, The annular air gap size of 6.05 mm is selected not to bring about too much hydraulic loss. Resultantly design analysis makes pump have the electrical input of 604 VA and the hydrodynamical capacity of 1.3 bars and 40 l/min.

• Nuclear Engineering and Technology
• /
• 제47권3호
• /
• pp.253-259
• /
• 2015

The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ~265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

• 한국전기전자재료학회:학술대회논문집
• /
• 한국전기전자재료학회 1999년도 춘계학술대회 논문집
• /
• pp.434-437
• /
• 1999

An EM pump is used for the purpose of transporting the electrically conducting liquid sodium of the high temperature that is used as a coolant in the liquid metal reactor. In the present study, the pilot pump has been designed and manufactured for the high temperature of $600^{\circ}C$ by the equivalent circuit materials and the consideration of the materials and functions. The length and diameter of the pump are given as 84 cm and 10 cm each due to the fixed geometry of the circulation system to be installed. The characteristic of the developing pressure and efficiency is found out by using Laithewaite\`s standard design formula. It is shown that the developing pressure and efficiency are maximized at the frequency of 15 Hz from the curve. The annular channel gap of 3.95 mm is selected in the range of the reasonable hydraulic frictional loss. The components of the pump consist of the material for the high temperature. And then, the pump is manufactured to have the nominal flowrate of 40 1/min and developing Pressure of 1.3 bar.

• Nuclear Engineering and Technology
• /
• 제51권3호
• /
• pp.654-664
• /
• 2019

The geometrical and electromagnetic variables of a rectangular-type magnetohydrodynamic (MHD) circulation system are optimized to solve MHD equations for the active decay heat removal system of a prototype Gen-IV sodium fast reactor. Decay heat must be actively removed from the reactor coolant to prevent the reactor system from exceeding its temperature limit. A rectangular-type MHD circulation system is adopted to remove this heat via an active system that produces developed pressure through the Lorentz force of the circulating sodium. Thus, the rectangular-type MHD circulation system for a circulating loop is modeled with the following specifications: a developed pressure of 2 kPa and flow rate of $0.02m^3/s$ at a temperature of 499 K. The MHD equations, which consist of momentum and Maxwell's equations, are solved to find the minimum input current satisfying the nominal developed pressure and flow rate according to the change of variables including the magnetic flux density and geometrical variables. The optimization shows that the rectangular-type MHD circulation system requires a current of 3976 A and a magnetic flux density of 0.037 T under the conditions of the active decay heat removal system.

• 한국압력기기공학회 논문집
• /
• 제20권1호
• /
• pp.75-83
• /
• 2024

In this study, high temperature design and integrity evaluation methodology have been developed for Type 316L stainless steel air-to-sodium heat exchanger which uses 700℃ sodium as coolant. Currently the only design rules that take creep effects into consideration explicitly for the 316L stainless steel subjected to high temperature in the creep range are French RCC-MRx, where elevated temperature designs are possible around 550℃. Absent design coefficients at high temperature were determined based on the material properties newly determined in previous studies, and high-temperature design evaluation methodologies were developed based on 3D finite element analyses on the 700℃ class heat exchanger. In addition, works were conducted on the web-based design evaluation program of HITEP_RCC-MRx including incorporation of material properties and design coefficients up to 700℃. Methodologies on high temperature design evaluations on Type 316L stainless steel high-temperature heat exchanger were suggested.

• Nuclear Engineering and Technology
• /
• 제54권2호
• /
• pp.589-600
• /
• 2022

Investigations on the molten-pool sloshing behavior are of essential value for improving nuclear safety evaluation of Core Disruptive Accidents (CDA) that would be possibly encountered for Sodium-cooled Fast Reactors (SFR). This paper is aimed at synthesizing the knowledge from our recent studies on molten-pool sloshing behavior with solid particles conducted at the Sun Yat-sen University. To better visualize and clarify the mechanism and characteristics of sloshing induced by local Fuel-Coolant Interaction (FCI), experiments were performed with various parameters by injecting nitrogen gas into a 2-dimensional liquid pool with accumulated solid particles. It was confirmed that under different particle-bed conditions, three representative flow regimes (i.e. the bubble-impulsion dominant, transitional and bed-inertia dominant regimes) are identifiable. Aimed at predicting the regime transitions during sloshing process, a predictive empirical model along with a regime map was proposed on the basis of experiments using single-sized spherical solid particles, and then was extended for covering more complex particle conditions (e.g. non-spherical, mixed-sized and mixed-density spherical particle conditions). To obtain more comprehensive understandings and verify the applicability and reliability of the predictive model under more realistic conditions (e.g. large-scale 3-dimensional condition), further experimental and modeling studies are also being prepared under other more complicated actual conditions.

• Corrosion Science and Technology
• /
• 제8권5호
• /
• pp.188-192
• /
• 2009

A forced outage due to a steam generator tube leak in a Korean nuclear power plant was reported.1) Primary water stress corrosion cracking has occurred in many tubes in the plant, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to understand the leak behavior of the tubes containing stress corrosion cracks. Stress corrosion cracks were developed in 0.1 M sodium tetrathionate solution at room temperature. Steam generator(SG) tubes with short cracks were successfully fabricated with a restricted solution contact method. The leak rates of the degraded tubes were measured at room temperature. Some tubes with 100 % through wall cracks showed an increase of leak rate with time at a constant pressure.