Safety approach of DFBR design study in Japan

The goal of the safety design for the demonstration fast breeder reactor is to ensure that the safety level is equivalent to or higher than that of the light water reactors of the same period. The design of the safety features such as reactor shutdown, decay heat removal and confinement systems is of importance to reach the goal. The reactor core is equipped with two independent fast shutdown systems, the primary system and the backup system. In addition, it is planned to strengthen the passive shutdown capability by using self- actuated systems such as a Curie point device for the backup system. The decay heat is removed from the core to the atmosphere through the safety lines of the direct reactor auxiliary cooling system which is composed of four independent lines. Furthermore, under the severe conditions that no active function of the decay heat removal system is available, the heat can be removed by natural convection through the safety lines by taking advantage of the high boiling temperature of sodium. For the confinement function, the reactor vessel is surrounded by a containment vessel and a confinement area.

The design concept of these safety features is described in this paper.     

Passive cooldown performance of a 65 MW integral reactor

Advanced integral-type pressurized water reactor with a maximum thermal power of 65 MW is under development at the Korea Atomic Energy Research Institute (KAERI). This 65 MW integral reactor incorporates a number of innovative design features. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the TASS/SMR code and a sensitivity analysis has been performed to evaluate a passive cooldown capability for various system conditions such as natural and forced circulation conditions for the reactor coolant system or the passive residual heat removal system, and number of active PRHRS trains. The reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation and the 65 MW integral reactor can cool the coolant to the SCS entry condition in the primary system for all the possible operational conditions.     

脉冲堆余热导出安全性实验研究

实验研究了脉冲堆余热导出的安全性,给出了停堆后燃料芯体温度和堆水池散热能力随时间变化的实验数据及其分析方法和结果.实验及其分析结果表明,脉冲反应堆余热导出是安全的.     

Analysis of decay heat removal from RBMK-1500 reactor in decommissioning phase by natural circulation of water and air

The main problem in nuclear energy is providing of safety at all stages of lifetime of nuclear installations in conditions of normal operation, accidents and at shutdown. Ignalina NPP, located in Lithuania, is one of the latest with RBMK reactors at highest capacity. Ignalina NPP has two units, both are closed for decommissioning now (in 2004 and 2009). Both units are equipped with RBMK-1500 reactors, the thermal power output is 4200 MW, the electrical power capacity is 1500 MW for each. In RBMK-1500 reactor the fuel assemblies remain for long time inside reactor core after the final shutdown. The paper discusses possibility of heat removal from the RBMK-1500 core at shutdown condition by natural circulation of water (1) and air (2) inside the fuel channels. In first case the decay heat from fuel assemblies is removed due to natural circulation of water and the piping above reactor core should be cooled by means of ventilation in the drum separator compartments. To warrant free access of air in to fuel channels (in the second case) the reactor cooling system should be completely dry out and the pressure headers and the steam discharge valves in steam lines should be opened. If mentioned conditions will be fulfilled, the reactor core will be cooled by natural circulation of water or air and fuel rods remain intact.     

中压安注对RRA连接模式下严重事故进程影响分析

针对中国改进型百万千瓦级压水堆（CPR1000）核电机组在中间停堆反应堆余热排出系统（RRA）连接模式下失去高低压安注和喷淋的冷却剂丧失事故（LOCA）,采用MAAP5程序对参考机组的反应堆堆芯、反应堆冷却剂系统以及安全壳系统进行模拟计算,同时结合计算结果分析中压安注系统对该严重事故序列进程的影响,并研究其对事故的缓解作用。分析结果表明,在RRA连接模式下出现LOCA导致的堆芯裸露和升温过程中,中压安注的及时注入能有效地限制堆芯的升温行为,并可对严重事故进程起到重要的缓解作用,甚至为事故工况下失去高低压安注和喷淋时避免堆芯完整性遭到破坏提供可能。最后,根据分析结果针对现行核电机组的运行规程提出改进建议:对于中压安注箱的行政隔离行为,只对其电气开关做相应的隔离操作,而对安全壳厂房内的阀门就地部分做挂牌警示,不做现场挂锁的操作,这样不仅可避免在正常运行工况下中压安注箱误注入行为的发生,同时能够在RRA连接模式下发生LOCA时有效地保障堆芯的完整性,在保证电厂正常安全运行的同时,提高了机组在该模式下发生严重事故的缓解能力。      

Safety features of the gas-cooled fast breeder reactor

The safety features of the gas-cooled fast breeder reactor (GCFR) are described in the context of the 300-MW(e) demonstration plant design. They are of two general types, inherent and design-related. The inherent features are principally associated with the helium coolant and the nuclear coefficients. Design-related features influencing safety include shutdown systems, residual heat removal systems, method of core support, and the prestressed concrete reactor vessel (PCRV). This paper discusses the safety-related aspects of each of these. Recently completed residual heat removal system reliability studies are also discussed. The probability of residual heat removal system failure in the GCFR is found to be lower than that described for light water reactors. The safety characteristics of larger plants are examined, and increases in size are found to improve GCFR safety margins.     

Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300 MWe PWR NPP

Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure.     

The experimental breeder reactor II inherent shutdown and heat removal tests — results and analysis

A test program is being conducted to demonstrate that a power-producing liquid-metal reactor (LMR) can (1) passively remove shutdown heat by natural convection, (2) passively reduce power in response to a loss of reactor flow, and (3) passively reduce power in response to a loss of the balance-of-plant heat sink. Measurements and pretest predictions confirm that natural convection is a reliable, predictable method of shutdown heat removal and suggest that safety-related pumps or pony motors are not necessary for safe shutdown heat removal in an LMR. Measurements from tests in which reactor flow and heat rejection to the balance of plant were perturbed show that reactivity feedbacks can passively control power and temperature. Data from these tests form a basis for additional tests including a complete loss of flow without scram and a complete loss of heat sink without scram.     

Above-ground HTGR design for passive decay heat removal

The design of a small high-temperature gas-cooled reactor (HTGR) for passive decay heat removal which could be located deeply underground was proposed previously. In the present work, analogue design analyses of passive decay heat removal for an above-ground long-life small prismatic HTGR was carried out to obtain the conditions for successful decay heat removal by radiation and conduction inside the reactor building, and by radiation and natural cooling by air at the outer surface of the reactor building. Sensitivity analysis of the peak temperatures of both the core and the reactor building after reactor shutdown was performed by changing the physical characteristics of the reactor regions. Enlarging the reactor building was found to be an effective way to reduce the peak reactor building temperature to within its design limit. By using the obtained condition for design parameters, the appropriate sizes of reactor core and reactor building were evaluated for some reactors. Consequently, criticality and burnup analyses for the proposed reactors were performed to confirm the possibility of designing a long-life core for the core size and reactor power which meet the condition of removing decay heat successfully. Using our design, all the reactors with 20 wt% uranium enrichment could be critical for over nine years.     

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。     

球床式高温气冷堆的余热不确定性分析

反应堆在停堆后相当长时间内仍具有较高的剩余发热是核电站的重要特性,也是核电站安全分析的关键。因此,对反应堆余热及其不确定性进行分析,对于合理设计余热排出系统、研究论证燃料元件在事故后的安全特性等均具有重要意义。本工作结合德国针对球床式高温气冷堆制定的余热计算标准,介绍了球床式高温气冷堆剩余发热及其不确定性的计算方法,并结合200 MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步物理设计,对长期运行在满功率平衡堆芯状态下的反应堆停堆后的余热及其不确定性进行了计算分析,为进一步的事故分析提供依据。     

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.     

中国先进研究堆停堆冷却措施及其主要特点

停堆后冷却问题是中国先进研究堆(CARR)重要的安全问题之一.冷却措施的实施对CARR的安全和建设投资有较重要的影响.CARR采用停堆初期的强迫循环及停堆后期全堆芯自然循环相结合的策略实现正常停堆和事故停堆后的堆芯冷却.停堆冷却的过程具体分为主泵大质量惯性飞轮惰转强迫冷却、应急堆芯冷却系统强迫冷却、自然循环功能部件动作实现全堆芯自然循环3个阶段.3个阶段既相互衔接又相互独立,每个阶段各有特点.停堆冷却策略的实施证明,CARR停堆冷却过程是可靠、有效、合理的,符合先进研究堆的发展趋势.     

Evaluation of decay heating in shutdown reactors

Heating from the decay of radioactive nuclides in shutdown reactors plays an important role in the safety evaluation of nuclear power plants. It also must be known in order to design spent fuel storage systems, shutdown reactor cooling systems and heat sink, reprocessing and nuclear waste disposal systems. Of these applications, the analysis of reactor accident scenarios has been the main impetus to develop more accurate methods of decay heat evaluation. It was recognized early in the 1970s that the knowledge was inadequate for safety requirements and that this placed an economic burden on nuclear power plants.Intensive research has been undertaken in the past few years and this has led to a much more precise knowledge of decay heat power in Light Water Reactors (LWR). With additional work this improvement can soon be extended to other reactors types. This paper reviews the background, recent research developments and the evolution of a major revision of the American Nuclear Society Standard for decay heat power in LWR.     

After heat removal by a closed secondary cooler circuit on the example of the advanced high temperature reactor AHTR 500 with central graphite column

The application of natural convection in connection with an after heat removal concept in general supports the claim for an inherent safety concept for advanced high temperature reactors (HTR). The effectivity of such an after heat removal (AHR) concept will be explored exemplarily for the process-heat reactor AHTR 500 with central graphite column by a thermohydraulic simulation of a secondary cooler circuit which is thermally connected with the primary circuit. This coupling is undertaken by an AHR-cooler located in the upper part of the graphite column. The heat removal from the secondary circuit is taking place outside of the reactor by a secondary heat exchanger under the assumption that the latter is cooled by a water capacity flow on an ambient temperature level. The developed calculation model determines iteratively the dynamic and thermal positions of equilibrium in the primary and secondary circuit which in the after heat removal mode of operation are exclusively run by natural convection. Different types of design for the central column heat exchanger (coaxial tube, U-tube and helically coiled tube heat exchanger) have been compared. For the secondary heat exchanger a parallel tube design has been supposed. The choice of the secondary flow medium as well as the most important limiting quantities influencing the transmission of heat via the secondary circuit during the after heat removal mode of operation are subject of a parameter study.     

Accident management following loss-of-coolant accidents during cooldown in a Westinghouse two-loop PWR

Operation of pressurised water reactors involves shutdown periods for refuelling and maintenance. In preparation for this, the reactor system is cooled down, depressurised and partially drained. Although reactor coolant pressure is lower than during full-power operation, there remains the possibility of a loss-of-coolant accident (LOCA), with a certain but low probability. While the decay heat to be removed is lower than that from a LOCA at full power, the reduced availability of safety systems implies a risk of failing to maintain core cooling, and hence of core damage. This is recognised though probabilistic safety analyses (PSA), which identify low but non-negligible contributions to core damage frequency from accidents during cooldown and shutdown. Analyses are made for a typical two-loop Westinghouse PWR of the consequences of a range of LOCAs during hot and intermediate shutdown, 4 and 5 h after reactor shutdown respectively. The accumulators are isolated, while power to some of the pumped safety injection systems (SIs) is racked out. The study assesses the effectiveness of the nominally assumed SIs in restoring coolant inventory and preventing core damage, and the margin against core damage where their actuation is delayed. The calculations use the engineering-level MELCOR1.8.5 code, supplemented by the SCDAPSIM and SCDAP/RELAP5 codes, which provide a more detailed treatment of coolant system thermal hydraulics and core behaviour. Both treatments show that the core is readily quenched, without damage, by the nominal SI which assumes operation of only one pump. Margins against additional scenario and model uncertainties are assessed by assuming a delay of 900 s (the time needed to actuate the remaining pumps) and a variety of assumptions regarding models and the number of pumps available in conjunction with both MELCOR and versions of SCDAP. Overall, the study provides confidence in the inherent robustness of the plant design with respect to LOCA during cooldown to cold shutdown, and in the validity of a two-tier calculational method. The results have been directly used in updating the plant shutdown PSA, by changing the success criteria for core cooling during cooldown of the plant and showing a reduction in overall risk.     

SADHANA facility for simulation of natural convection in the SGDHR system of PFBR

The Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled pool type fast reactor being constructed at Kalpakkam, India. PFBR has all the reactor components immersed in the pool of sodium and the fission heat generated in the core, is removed by the sodium circulating in the pool. During normal operation this fission heat is transferred by primary sodium to secondary sodium, which in turn transfers the heat to water in the steam generator for producing steam. The removal of the decay heat generated in the reactor core after the reactor shutdown is also very important to maintain the structural integrity of reactor core components. PFBR employs two independent systems namely, Operational Grade Decay Heat Removal system (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS) for decay heat removal. SGDHR system is a passive system working on natural convection to ensure the core coolability even under station blackout condition. It is very important to study the thermal hydraulic behavior of Safety Grade Decay Heat Removal system of PFBR to ensure its reliable operation. A scaled down model of the circuit, named SADHANA has been modeled, designed, constructed and commissioned for demonstration and evaluation of these systems. The facility has completed around 2000 h of high temperature operation. The performance of the experimental system is satisfactory and it meets all the design requirements. At 550 °C sodium pool temperature in test vessel the secondary sodium loop generated a sodium flow of 6.7 m³/h. These experiments have revealed the adequacy and capability of SGDHR system to remove the decay heat from the fast breeder reactor core after its shutdown.     

CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal–hydraulic analyzes of gas cooled reactors

The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal–hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coeffcient, …), critical position of control rods, reactivity insertion aspects, …. For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, …) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal–hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark problems such as the HTR-10 thermal–hydraulic exercise. Examples of containment thermal–hydraulics calculations for fast reactor design (GFR) are also detailed.     

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.     

Experimental study on the natural circulation of a passive residual heat removal system for an integral reactor following a safety related event

Natural circulation characteristics of an integral type reactor during the operation of a passive residual heat removal system (PRHRS) following a safety related event has been experimentally investigated by using the VISTA facility. A PRHRS actuation trip signal is generated by a high power trip signal following a steam flow increasing event. The experimental results show that the single-phase coolant flows steadily in the primary loop by a natural convection process and that it effectively removes the decay heat from the core through a steam generator during the PRHRS operation. The heat transfers through the PRHRS heat exchanger and the emergency cooldown tank (ECT) are sufficient enough to enable a two-phase natural circulation of the coolant in the PRHRS loop.