• Nuclear Engineering and Technology
• /
• 제52권12호
• /
• pp.2789-2802
• /
• 2020

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

• Nuclear Engineering and Technology
• /
• 제51권7호
• /
• pp.1738-1748
• /
• 2019

Previous research has shown that the inclined condition has an impact on the natural circulation (natural circulation) mode operation of Floating Nuclear Power Plant (FNPP) mounted on the movable marine platform. Due to its compact structure, small volume, strong maneuverability, the Integral Pressurized Water Reactor (IPWR) is adopted as marine reactor in general. The OTSGs of IPWR are symmetrically arranged in the annular region between the reactor vessel and core support barrel in this paper. Therefore, many parallel natural circulation loops are built between the core and the OTSGs primary side when the main pump is stopped. and the inclined condition would lead to discrepancies of the natural circulation drive head among the OTSGs in different locations. In addition, the flow rate and temperature nonuniform distribution of the core caused by inclined condition are coupled with the thermal hydraulics parameters maldistribution caused by OTSG group operating mode on low power operation. By means of the RELAP5 codes were modified by adding module calculating the effect of inclined, heaving and rolling condition, the simulation model of IPWR in inclined condition was built. Using the models developed, the influences on natural circulation operation by inclined angle and OTSG position, the transitions between forced circulation (forced circulation) and natural circulation and the effect on natural circulation operation by different OTSG grouping situations in inclined condition were analyzed. It was observed that a larger inclined angle results the temperature of the core outlet is too high and the OTSG superheat steam is insufficient in natural circulation mode operation. In general, the inclined angle is smaller unless the hull is destroyed seriously or the platform overturn in the ocean. In consequence, the results indicated that the IPWR in the movable marine platform in natural circulation mode operation is safety. Selecting an appropriate average temperature setting value or operating the uplifted OTSG group individually is able to reduce the influence on natural circulation flow of IPWR by inclined condition.

• Nuclear Engineering and Technology
• /
• 제33권6호
• /
• pp.638-651
• /
• 2001

The present study is to assess the applicability of the best-estimate thermal-hydraulic code, MARS 1.4, for the analysis of thermal-hydraulic behavior in PWRs during natural circulation conditions. The code simulates a natural circulation test, BETHSY test 4. la, which was conducted on the integral test facility of BETHSY. The test represented the cooling states of the primary cooling system under single-phase natural circulation, two-phase natural circulation and the reflux condensation mode with conditions corresponding to the residual power, 2% of the rated core power value and 6.8 MPa at the secondary system. Based on MARS 1.4 calculations, the major thermal-hydraulic behaviors during natural circulation are evaluated and the differences between the experimental data and calculated results are identified. The calculated results show generally good behavior with regard to the experimental results; the region of single-phase natural circulation is 100-92% of the initial mass inventory, two-phase natural circulation is 84-63 %, and the reflux condensation mode occurred below 58 %. U-tubes empty and the core uncovery are obtained at 39 % and 34 % of the initial mass inventory, respectively.

• Nuclear Engineering and Technology
• /
• 제54권3호
• /
• pp.826-833
• /
• 2022

A two-phase natural circulation test using SMART integral test loop (SMART-ITL) was conducted to explore thermo-hydraulic phenomena of two-phase natural circulation in the SMART reactor. Specifically, the test examined the natural circulation in the primary loop under a stepwise coolant inventory loss while keeping the core power constant at 5% of the scaled full power. Based on the test results, three flow regimes were observed: single-phase natural circulation (SPNC), two-phase natural circulation (TPNC), and boiler-condenser natural circulation (BCNC). The flow rate remained steady in the SPNC, slightly increased in the TPNC, and dropped abruptly and maintained in the BCNC. Using a natural circulation flow map, the natural circulation characteristic in the SMART-ITL was compared with those in pressurized water reactor simulators. In the SMART-ITL, a BCNC regime appeared instead of siphon condensation and reflux condensation regimes because of the use of once-through steam generators.

• Nuclear Engineering and Technology
• /
• 제54권2호
• /
• pp.554-564
• /
• 2022

The natural circulation phenomena occurring in fully integrated nuclear reactors are associated with a unique formation mechanism. The phenomenon results from a structural feature of these reactors involving upward flow from the core, located in the central-bottom region of a single vessel, and downward flow to the steam generator in the annulus region. In this study, to understand the natural circulation in a single vessel involving a multi-layered flow path, single-phase and two-phase natural circulation tests were performed using the SMART-ITL facility, and validation analysis of the TASS/SMR-S code was performed by comparing the corresponding test results. Three single-phase natural circulation tests were sequentially conducted at 15%, 10%, and 5% of full-scaled core-power without RCP operation, following which a two-phase natural circulation test was successively conducted with an artificial discharge of coolant inventory. The simulation capability of the TASS/SMR-S code with respect to the natural circulation phenomena was validated against the test results, and somewhat conservative but reasonably comparative results in terms of overall thermalhydraulic behavior were shown.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
• /
• pp.458-463
• /
• 1997

A transient natural circulation behavior during a LOCA at hot-standby operation is evaluated for YGN Units 3/4. The plant initial condition is determined within the EOP limitation as suitable to hot-standby mode and the transient scenario is prepared as relevant to evaluation of transient natural circulation. A 0.4% cold leg break with loss of off-site power is calculated with RELAP5/MOD3.2, whose predictability has been verified for SBLOCA natural circulation test, S-NC-8B. Through one hour transient analysis, it is found that the plant has its own decay heat removal capability by natural circulation following a LOCA, at hot-standby mode. Additional calculation is performed to investigate an effect of HPSI flow on natural circulation.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2004년도 춘계학술대회
• /
• pp.1927-1932
• /
• 2004

A 1/21.6 scaled non-heating experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The behaviors of the air bubble-induced two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the injected air flow rate and distribution. As the injected air flow rates increased, the natural circulation flow rates also increased. Both the longitudinal and the latitudinal distributions of the injected air affected the natural circulation flow rates, especially, the longitudinal effect is more larger.

• 대한기계학회:학술대회논문집
• /
• 대한기계학회 2007년도 춘계학술대회B
• /
• pp.3086-3091
• /
• 2007

When a molten corium is relocated in a lower head of a reactor vessel, the ERVC (External Reactor Vessel Cooling) system is actuated as coolant is supplied into a reactor cavity to remove a decay heat from the molten corium during a severe accident. To achieve this severe accident mitigation strategy, the two-phase natural circulation flow in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. For this reason, one-dimensional natural circulation flow tests were conducted to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled-down as the half height and 1/238 rectangular channel area of the APR1400 reactor vessel. As the water inlet area increased, the natural circulation mass flow rate asymptotically increased, that is, it converged at a specific value. And the circulation mass flow rate also increased as the outlet area, injected air flow rate, and outlet height increased. But the circulation mass flow rate was not changed along with the external water level variation if the water level was higher than the outlet height.

• 대한기계학회논문집B
• /
• 제22권12호
• /
• pp.1763-1773
• /
• 1998

In the present experimental study, the flow behavior in a semi-closed two-phase natural circulation loop was examined. Water was used as the working fluid. Heat flux, heater-inlet subcooling, and flow restrictions at the heater-inlet and at the expansion-tank-line were taken as the controlling parameters Six circulation modes were identified by changing heat flux and inlet subcooling conditions ; single-phase continuous circulation, periodic circulation (A), two-phase continuous circulation, and periodic circulations (B), (C), and (D). Among these, the single-phase and two-phase continuous-circulation modes exhibit no significant oscillations and are considered to be stable. Periodic circulation (A) is characterized by the large amplitude two-phase f10w oscillations with the temporal single-phase circulation between them, while periodic circulation (B) featured by the flow oscillations with continuous boiling inside the heater section. Periodic circulation (C) appears to be the manometric oscillation with continuous boiling. Periodic circulation (D) has the longer period than periodic circulation (B) and a substantial amount of liquid flow back and forth through the expansion-tank-line periodically ; this mode is considered the pressure drop oscillation. Parametric study shows that the increases of the inlet- and expansion-tank-line- restrictions and the decrease of inlet subcooling broaden the range of the stable two-phase(continuous circulation) mode.

• 한국추진공학회지
• /
• 제23권5호
• /
• pp.43-52
• /
• 2019

극저온 유체를 사용하는 발사체는 극저온 유체의 자연순환회로를 이용하여 발사체의 엔진 입구를 냉각한다. 자연순환회로의 질량유량은 순환시스템을 구성하는 배관의 길이 및 직경과 시스템으로 들어오는 열유입에 의하여 결정된다. 극저온 유체의 자연순환회로의 순환 검증 및 질량유량 측정을 위하여 실험을 진행하였으며, 이론적 계산 결과와 비교하였다. 비교 결과 12%의 오차가 있음을 확인하였다. 이 결과를 바탕으로 발사체 상단에서 저중력 구간 및 가속 구간에서의 자연순환 질량유량을 예측한 내용을 포함한다. 가속구간에서는 산화제탱크가 100 kPa 내외로 유지하는 것이 자연순환유량 증가에 이로웠으며, 저중력구간에서는 중력가속도의 크기에 따른 최적 압력으로 조절해야 자연순환유량의 최고값을 유지할 수 있었다.