Implications of the EBR-II inherent safety demonstration test

Extensive thermal-hydraulics testing at EBR-II culminated in the Inherent Safety Demonstration Test on April 3, 1986. This work may well lead to fundamental changes in the approach to the design and licensing of liquid-metal-cooled reactor (LMR) power plants. The EBR-II test program has thus far demonstrated (1) passive removal of decay heat by natural circulation, (2) passive reactor shutdown for a loss of flow without scram, and (3) passive reactor shutdown for a loss of heat sink without scram. Supporting analyses indicate that these characteristics can be incorporated into larger commercial LMRs and be used as the basis for a totally new passive control strategy. Analyses and tests are now in progress to show that LMRs with these characteristics and the passive control strategy are also inherently safe for unprotected overpower accidents.     

Performance evaluation of an advanced integral reactor against an anticipated transient without scram

Performance evaluation of KAERI’s advanced integral reactor against an anticipated transient without scram has been carried out with the transients and setpoint simulation/small and medium reactor code, by considering a decrease in the heat removal by the secondary system, a loss of offsite power and an inadvertent control rod withdrawal event as an initiating event. In a decrease in the heat transfer by the secondary system and a loss of offsite power, the reactor coolant system pressures can be maintained below 110% of the design pressure during the transition period due to the effect of the large negative moderator temperature coefficient. On the other hand, in an inadvertent control rod withdrawal event, the pressure of the reactor coolant system increases up to the ASME service level C stress limit due to a high reactivity insertion into a reactor core by the adoption of a boron free core concept. Therefore, a hardware installation against an anticipated transient without scram is essential to mitigate the consequences resulting from an inadvertent control rod withdrawal event. A diverse protection system, which is an independent and diverse reactor shutdown system that is initiated by the signals of a high core power or a high pressurizer pressure, is adopted in the advanced integral reactor. According to the reassessment results by considering the diverse protection system for a reactor shutdown, the diverse protection system is helpful in mitigating the consequences of an anticipated transient without scram.     

Simulation of two regime sodium flows in a parallel plate channel during buoyancy-driven flow

Due to the remaining decay heat, the reactor core has yet to be cooled after shutdown. As the reactor power is low, the core can be sufficiently cooled by natural convection. The coolant flow is driven by buoyancy, as the heated fluid decreases its density. During buoyancy-driven flow, a reverse flow may take place when a heat sink exists close to the heat source, such as in a wall (edge) or corner subchannels. For simplicity in applying boundary condition, the reverse flow is simulated by two parallel plates, one as a heat source having positive heat flux and the other as heat sink with negative heat flux.     

Lead-cooled flexible conversion ratio fast reactor

Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO₂ (S-CO₂) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO₂ PCS.     

Evolution of thermal-hydraulics testing in EBR-II

A thermal-hydraulics testing and modeling program has been underway at the Experimental Breeder Reactor-II (EBR-II) for 12 years. This work culminated in two tests of historical importance to commercial nuclear power, a loss of flow without scram and a loss of heat sink without scram, both from 100% initial power. These tests showed that natural processes will shut EBR-II down and maintain cooling without automatic control rod action or operator intervention. Supporting analyses indicate that these results are characteristic of a range of sizes of liquid metal cooled reactors (LMRs), if these reactors use metal driver fuel. This type of fuel is being developed as part of the Integral Fast Reactor Program at Argonne National Laboratory. Work is now underway at EBR-II to exploit the inherent safety of metal-fueled LMRs with regard to development of improved plant control strategies.     

Analysis of loss of coolant accident in MTR pool type research reactor

In MTR research reactors, heat removal is, safely performed by forced convection during normal operation and by natural convection after reactor shutdown for residual decay heat removal. However, according to the duration time of operation at full power, it may be required to maintain the forced convection, for a certain period of time after the reactor shutdown. This is among the general requirements for the overall safety engineering features of MTR research reactors to ensure a safe residual heat removal. For instance, in safety analysis of research reactors, initiating events that may challenge the safe removal of residual heat must be identified and analyzed.In the present work, it was assumed a total loss of coolant accident in a typical MTR nuclear research reactor with the objective of examining the core behavior and the occurrence of any fuel damage.For this purpose, the IAEA 10 MW benchmark core, which is a representative of medium power pool type MTR research reactors, was chosen herein in order to investigate the evolution of cladding temperature through the use of a best estimate thermalhydraulic system code RELAP5/mod3.2.     

Safety analysis of small long life CANDLE fast reactor

Safety analysis for small long life fast CANDLE reactor was performed with ULOF (unprotected loss of flow), SDRW (unprotected shut down rods withdrawal), ULOHS (unprotected loss of heat sink) and LB (local blockage) accidents. The employed reactor system is based on the former steady state research. The core with 1.0 m radius and 2.0 m length produces 200 MW thermal power in steady state, using enriched N-15 natural uranium as fresh fuel and lead bismuth as coolant. The former 3 accidents were simulated without scram by neutronic-thermal hydraulic calculation coupled with stationary diffusion calculation. The LB accident was simulated by transient thermal hydraulic calculation only, because in this accident the neutronic factors basically do not change. The analysis results show that the proposed small CANDLE fast reactor can survive all the accidents without any active protection.     

EBR-II unprotected loss-of-heat-sink predictions and preliminary test results

Two unprotected (i.e., no scram or plant protection system action) loss-of-heat-sink transients were performed on the Experimental Breeder Reactor-II in the Spring of 1986. One was initiated from full power (60 MW) and the other from half power. The loss of heat sink was accomplished in each test by essentially stopping the secondary-loop sodium coolant flow. Pretest predictions along with preliminary test results demonstrate that the reactor shuts itself down in a benign and predictable manner in which all of the reactor temperatures approach a quenching (or smothering) temperature at which the fission power goes to zero.     

Loss-of-primary-flow-without-scram tests: Pretest predictions and preliminary results

A series of tests in the Experimental Breeder Reactor No. 2 (EBR-II) has been concluded that investigated the effects of a complete loss of primary flow without scram. The development and preliminary study of these events is first discussed, including the test limits and controlling parameters. The results of two of the tests, SHRT 39 and 45, are examined in detail, although a compact summary of all the tests is included. The success in meeting the objectives of the test program served to verify that natural processes will shut down the reactor and maintain adequate cooling without control rod or operator intervention. The good comparison between predicted and measured results confirms that such events can be analyzed without elaborate codes if the basic processes are understood. Furthermore, recent studies suggest that the EBR-II results are characteristic of new innovative LMR designs being pursued in the U.S. that incorporate metallic driver fuel.     

Stability analysis of a liquid-metal reactor and its primary heat transport system

Part of the reactor design process is the assessment of the impact of different design changes on pre-defined performance criteria including stability of the reactor system under different conditions. This work focuses on the stability analysis of a combined liquid-metal reactor and primary heat transport system where system parameters are free to vary, with particular interest in low reactor power, low reactor coolant flow conditions. Such conditions might be encountered, for example, after a loss of flow without scram in some passively safe reactor designs. Linear-stability-analysis-based methods are developed to find the stability regions, stability boundary surface in system parameter space, and frequency of oscillation at oscillatory instability boundaries. Models are developed for the reactor, detailed thermal hydraulic reactivity feedback associated with coolant outlet and inlet temperatures, decay heat and primary system. The developed stability analysis tools are applied to the system model. The system parameters include integral reactivity parameters, decay heat, primary system mass, coolant flow and natural circulation flow. The resulting stability boundary surface and its associated frequency of oscillation surface in multidimensional system parameter space show the effect of system parameter changes. By adopting model parameters from liquid-metal reactor designs, a stability prediction procedure is illustrated.     

Thermal hydraulic challenges of Gas Cooled Fast Reactors with passive safety features

Transient response of a Gas Cooled Fast Reactor (GFR) coupled to a recompression supercritical CO₂ (S-CO₂) power conversion system (PCS) in a direct cycle to a Loss of Coolant Accident (LOCA) and a Loss of Generator Load Accident is analyzed using RELAP5-3D. A number of thermal hydraulic challenges for GFR design are pointed out as the designers strive to accommodate cooling of the high power density core of a fast reactor by a gas with its inherently low heat transfer capability, in particular under post-LOCA events when system pressure is lost and when reliance on passive decay heat removal (DHR) is emphasized. Although it is possible to design a S-CO₂ cooled GFR that can survive LOCA by cooling the core through natural circulating loops between the core and elevated emergency cooling heat exchangers, it is not an attractive approach because of various bypass paths that can, depending on break location, degrade core cooling. Moreover, natural circulation gas loops can operate in deteriorated heat transfer regimes with substantial reduction of heat transfer coefficient: as low as 30% of forced convection values, and data and correlations in these regimes carry large uncertainties. Therefore, reliable battery powered blowers for post-LOCA decay heat removal that provide flow in well defined regimes with low uncertainty, and can be easily overdesigned to accommodate bypass flows were selected. The results confirm that a GFR with such a DHR system and negative coolant void worth can withstand LOCA with and without scram as well as loss of electrical load without exceeding core temperature and turbomachinery overspeed limits.     

Accident analysis of lead or lead—bismuth cooled small safe long-life fast reactor using metallic or nitride fuel

Safety analysis of a lead or lead—bismuth cooled small safe long-life fast reactor was performed. It is proposed that the reactor be used in relatively isolated areas, and operated to the end of its life without refueling or fuel shuffling. In the present paper the reactor power and lifetime are set at 150 MWt and 12 years respectively. In order to assume its safety performance, the following accidents without scram were simulated with neutronic-thermal-hydraulic analysis: unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), simultaneous ULOF and UTOP accidents, and simultaneous ULOF, UTOP and unprotected loss of heat sink (ULOHS) accidents. For each type of accident, four types of long-life small reactor (lead cooled metallic fueled, lead cooled nitride fueled, lead-bismuth cooled metallic fueled, and lead—bismuth cooled nitride fueled) were analyzed. It is shown that all the proposed designs can survive these accidents without requiring help from the operator or active devices.     

Reactivity insertion limits in a typical pool-type research reactor cooled by natural circulation

The course of reactivity insertion in a pool type research reactor, with scram disabled under natural circulation condition is numerically investigated. The analyses were performed by a coupled kinetic–thermal–hydraulic computer code developed specifically for this task. The 10-MW IAEA MTR research reactor was subjected to unprotected reactivity insertion (step and ramp) for both low and high-enriched fuel with continuous reactivity feedback due to coolant and fuel temperature effects. In general, it was found that the power, core mass flow rate and clad temperature under fully established natural circulation are higher for high-enriched fuel than for low enriched fuel. This is unlike the case of decay heat removal, where equal clad temperatures are reported for both fuels. The analysis of reactivity represented by the maximum insertion of positive reactivity ($0.73) demonstrated the high inherent safety features of MTR-type research reactor. Even in the case of total excess reactivity without scram, the high reactivity feedbacks of fuel and moderator temperatures limit the power excursion and avoid consequently escalation of clad temperature to the level of onset of nucleate boiling and sub-cooled void formation. The code can also be modified to provide an accurate capability for the analyses of research reactor transients under forced convection.     

Increasing the power of the high temperature reactor module

To alleviate the economic problems of the modular pebble bed high temperature reactor, its design was modified in such a way that the power output was increased from 200 to 350 MWth. The core geometry was changed from cylindrical to annular, and the pressure vessel diameter was increased to 6.7 m. Control rods are placed in both the outer reflector and the graphite central column. In a safety analysis, loss of heat sink, loss of coolant and water ingress accident were examined. Reactor shutdown and decay heat removal take place passively, and the maximum fuel temperature stays theoreticallybelow 1600 °C, implying full retention of the fission products in the fuel elements. The central column has a diminishing effect on the positive reactivity effect of water ingress. A cost analysis shows that the specific investment costs of a four-module plant would decrease by 26% and the electricity generating costs would reduce by 19%.     

EBR-Ⅱ余热排出实验及非能动余热排出系统性能分析

本文基于SAC-CFR事故分析程序,在国际原子能机构联合研究项目（IAEA CRP）框架下,对美国EBR-Ⅱ快堆余热排出实验（SHRT-17、SHRT-45R）进行了分析,计算了事故余热排出系统（DRACS）的响应、衰变热功率、关键部件的冷却剂温度、一回路的质量流量等关键参数。将计算参数与实验数据进行了对比,对程序的有效性进行了验证。计算结果表明,在SHRT-17工况下,随DRACS风门的打开,每台事故热交换器可带走330 406.4 W的堆芯余热,DRACS具有长期带走衰变热的能力。     

兆瓦级空间气冷堆系统启停堆瞬态分析

本文针对兆瓦级高温气冷堆布雷顿循环系统,采用Fortran语言开发系统分析程序TASS,包括堆芯、透平-发电机-压气机、回热器、冷却器和热管式辐射散热器等模型。通过设计值与程序计算值对比对TASS进行验证,并利用TASS对系统启动、停堆瞬态工况进行数值模拟。结果显示,通过分两阶段、阶梯式引入正反应性和提高涡轮机械的转轴速度,堆芯流量和功率匹配良好,系统可在3.5 h内完成启动过程,达到反应堆功率3 406 kW、流量14.2 kg/s的稳态运行。系统停堆过程中,反应堆可依靠自身的非能动余热排出能力,确保芯块和包壳温度与熔点间存在较大安全裕量,实现安全停堆。     

Natural circulation capability of Pb-Bi cooled fast reactor: PEACER

First-principle calculations were performed to analyze the natural circulation heat removal from the core of a liquid metal reactor (LMR). The lead-bismuth (Pb-Bi) was chosen as the primary coolant for the LMR system. From the single channel analysis the temperature and the pressure drop are calculated along the fuel assembly. The total pressure drop of the core is less than 100kPa due to the large pitch-to-diameter ratio and the small height of the fuel pin. The natural circulation potential is a key characteristics of the LMR design. The steady-state momentum and energy equations are solved along the primary coolant path. The calculations are divided into two parts: an analytical model and a one-dimensional lumped parameter flow loop model. Results of the analytical model indicate that the elevation difference of 4.5m between thermal centers of the core and the steam generators could remove as much as 10% of the nominal operating reactor power. The flow loop model yielded the total pressure drop and the natural circulation heat removal capacity.     

基于SAC-CFR系统分析软件的EBR-Ⅱ LOHSWS事故分析

利用自主开发的系统分析软件SAC-CFR对美国实验增殖堆2号(EBR-Ⅱ)的未能紧急停堆的丧失热阱(LOHSWS)事故全厂瞬态行为进行建模分析。SAC-CFR耦合了新开发的三维钠池计算模型,用于分析EBR-Ⅱ钠池内的流型。结果表明,SAC-CFR计算结果与实验数据相符合,SAC-CFR可用于快堆部分事故工况的瞬态计算,同时也证实了EBR-Ⅱ可在LOHSWS事故下依靠固有安全性停堆。     

An experimental study on natural circulation decay heat removal system for a loop type fast reactor

A fully natural circulation-based system is adopted in the decay heat removal system (DHRS) of an advanced loop type fast reactor. Decay heat removal by natural circulation is a significant passive safety measure against station blackout. As a representative of the advanced loop type fast reactor, DHRS of the sodium fast reactor of 1500 MWe being designed in Japan comprises a direct reactor auxiliary cooling system (DRACS), which has a dipped heat exchanger in the reactor vessel, and two units of primary reactor auxiliary cooling system (PRACS), which has a heat exchanger in the primary-side inlet plenum of an intermediate heat exchanger in each loop. The thermal-hydraulic phenomena in the plant under natural circulation conditions need to be understood for establishing a reliable natural circulation driven DHRS. In this study, sodium experiments were conducted using a plant dynamic test loop to understand the thermal-hydraulic phenomena considering natural circulation in the plant under a broad range of plant operation conditions. The sodium experiments simulating the scram transient confirmed that PRACS started up smoothly under natural circulation, and the simulated core was stably cooled after the scram. Moreover, they were conducted by varying the pressure loss coefficients of the loop as the experimental parameters. These experiments confirmed robustness of the PRACS, which the increasing of pressure loss coefficient did not affect the heat removal capacity very much due to the feedback effect of natural circulation.     

The effect of primary pump coastdown characteristics on loss-of-flow transients without scram in EBR-II

A series of loss-of-flow (LOF) tests without scram (unprotected) is planned for the Experimental Breeder Reactor II (EBR-II) to demonstrate the inherent shutdown capability of the reactor during an LOF event. The purpose of this paper is to discuss in detail the unprotected LOF transient analysis, the validation of the EBR-II reactivity feedback modeling, and the significance of pump coastdown characteristics on peak reactor temperatures. The tests as designed are limited by the fuel-cladding eutectic temperature of the fuel elements, and in order to meet the required temperature limit, the initial power and flow of all the tests are 16.7 and 20% of their rated values, respectively. To further reduce peak temperature, the primary tank temperature is to be decreased to 338°C from the nominal 371°C. The results show that primary flow coastdown rate and the capacity of the auxiliary pump have dramatic effects on the reactor temperatures. The impact of secondary flow depends somewhat on test conditions. When the auxiliary pump is in operation, the effect of secondary flow behavior on the reactor temperature becomes less significant during an unprotected LOF event.