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.     

Thermohydraulic transient studies of the Chinese 200 MWe HTR-PM for loss of forced cooling accidents

Pressurized and Depressurized Loss Of Forced Cooling (PLOFC and DLOFC) are two important design basis accidents for high temperature gas-cooled reactors. Analysis of the reactor characteristic behaviors during LOFC can provide useful reference to the physics, thermohydraulic and structure designs of the reactor core, and can also verify the design of the Residual Heat Removal System (RHRS). The 200 MWe High Temperature gas-cooled Reactor Pebble-bed Module project (HTR-PM), designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China, is characterized by its inherent safety features, such as shutdown ability via negative temperature coefficients of reactivity, passive mechanism of decay heat removal and so on.     

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.     

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.     

PASCAR: Long burning small modular reactor based on natural circulation

PASCAR is a 100 MWt/35 MWe lead-bismuth-cooled small modular reactor which requires no on-site refueling and well suits to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. This paper includes both steady-state and transient performance evaluations for neutronics and thermal-hydraulics. Through design optimization studies for minimizing a burn-up reactivity loss, the metallic fuels-loaded core was designed with less than 1$ reactivity swing over 20-year cycle. A radial peaking power location shows the slow inward migration from outer enrichment zones while maintaining peaking factor within 1.35, reducing radiation damage and corrosion duty of high temperature environments. Equipped with coolant flow path large enough to ensure low pressure drop, this reactor is intended to operate by only natural circulation of chemically inert coolant within relatively low temperature range, 320-420 °C. Peak outlet temperature is nearly 450 °C where an Al-containing duplex cladding has sufficient corrosion resistance. Despite of 50% decrease of fuel thermal conductivity after swelling, inherent negative reactivity feedback and passive decay heat removal capability could secure an ample safety margin of peak fuel centerline temperature in tow safety analyses, unprotected transient overpower and unprotected loss of heat sink. The likelihood of loss of coolant, loss of flow, and local blockage is virtually eliminated by employing respectively a double-walled vessel, pump-less cooling, and cross-flow allowed open square assemblies. Simple fabrication, modular construction, and long burning cycle would compensate for economic disadvantages over smaller power and lower temperature than those of conventional fast reactors.     

泳池式反应堆自启停控制技术研究

反应堆实现自动启停,可以有效减轻运行人员工作强度,减少误操作,提高反应堆启动运行的安全可靠性。本文基于对典型泳池式反应堆的工艺特点以及启动操作的分析,对泳池式反应堆自启停系统的控制范围、层次结构、断点、典型控制逻辑进行研究,并搭建泳池式反应堆自启停的仿真测试系统。该自启停系统能够实现泳池式反应堆的自动启停,启停过程无人工操作,降低人员误操作可能性。      

Thermal hydraulic design of a liquid salt-cooled flexible conversion ratio fast reactor

A 2400 MWth liquid-salt cooled flexible conversion ratio reactor was designed, utilizing the ternary chloride salt NaCl-KCl-MgCl₂ (30-20-50%) as coolant. The reference design uses a wire-wrapped, hexagonal lattice core, and is able to achieve a core power density of 130 kW/l with a core pressure drop of 700 kPa and a maximum cladding temperature under 650 °C. Four kidney-shaped conventional tube-in-shell heat exchangers are used to connect the primary system to a 545 °C supercritical CO₂ power conversion system. The core, intermediate heat exchangers, and reactor coolant pumps fit in a vessel approximately 10 m in diameter and less than 20 m high. Lithium expansion modules (LEMs) were used to reconcile conflicting thermal hydraulic and reactor physics requirements in the liquid salt core. Use of LEMs allowed the design of a very favorable reactivity response which greatly benefits transient mitigation. A reactor vessel auxiliary cooling system (RVACS) and four redundant passive secondary auxiliary cooling systems (PSACSs) are used to provide passive heat removal, and are able to successfully mitigate both the unprotected station blackout transient as well as protected transients in which a scram occurs. Additionally, it was determined that the power conversion system can be used to mitigate both a loss of flow accident and an unprotected transient overpower.     

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.     

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.     

钠冷快堆熔断式非能动停堆系统方案设计

为保证和增强池式快堆的安全性,通过对比分析现有的非能动停堆装置,基于将某些合金在特定温度下拉伸强度发生突变的特性作为钠冷快堆非能动停堆的触发条件,提出了一种钠冷快堆熔断式非能动停堆系统的设计概念,能在发生无保护超功率事故或无保护失流事故的情况下引入负反应性。针对中国实验快堆(CEFR)的设计完成了熔断式非能动停堆系统的方案设计论证,并利用分析程序DYN4G对这一非能动停堆系统在CEFR无保护事故下的响应情况进行了模拟计算,由此得到了其组件设计的关键参数。分析结果表明,通过合理设计,在发生无保护事故时,熔断式非能动停堆系统能有效降低事故情况下的堆芯燃料组件及冷却剂的温度,进一步提高了钠冷快堆应对严重事故的能力。     

Thermal response of a modular high temperature reactor during passive cooldown under pressurized and depressurized conditions

The concept of inherent safety features of the modular HTR design with respect to passive decay heat removal through conduction, radiation and natural convection was first introduced in the German HTR-module (pebble fuel) design and subsequently extended to other modular HTR design in recent years, e.g. PBMR (pebble fuel), GT-MHR (prismatic fuel) and the new generation reactor V/HTR (prismatic fuel).This paper presents the numerical simulations of the V/HTR using the thermal-hydraulic code THERMIX which was initially developed for the analysis of HTRs with pebble fuels, verified by experiments, subsequently adopted for applications in the HTRs with prismatic fuels and checked against the results of CRP-3 benchmark problem analyzed by various countries with diverse codes.In this paper, the thermal response of the V/HTR (operating inlet/outlet temperatures 490/1000 °C) during post shutdown passive cooling under pressurized and depressurized primary system conditions has been investigated. Additional investigations have also been carried out to determine the influence of other inlet/outlet operating temperatures (e.g. 490/850, 350/850 or 350/1000 °C) on the maximum fuel and pressure vessel temperature during depressurized cooldown condition. In addition, some sensitivity analyses have also been performed to evaluate the effect of varying the parameters, i.e. decay heat, graphite conductivity, surface emissivity, etc., on the maximum fuel and pressure vessel temperature. The results show that the nominal peak fuel temperatures remain below 1600 °C for all these cases, which is the limiting temperature relating to radioactivity release from the fuel. The analyses presented in this paper demonstrate that the code THERMIX can be successfully applied for the thermal calculation of HTRs with prismatic fuel. The results also provide some fundamental information for the design optimization of V/HTR with respect to its maximum thermal power, operating temperatures, etc.     

Comparative study of unprotected loss of flow accident analysis of 1000 MWe and 500 MWe Fast Breeder Reactor Metal (FBR-M) cores and their inherent safety

Unprotected loss of flow (ULOF) analysis of metal (U–Pu–6% Zr) fuelled 500 MWe and 1000 MWe pool type FBR are studied to verify the passive shutdown capability and its inherent safety parameters. Study is also made with uncertainties (typically 20%) on the sensitive feedback parameters such as core radial expansion feedback and sodium void reactivity effect. Inference of the study is, nominal transient behavior of both 500 MWe and 1000 MWe core are benign under unprotected loss of flow accident (ULOFA) and the transient power reduces to natural circulation based Safety Grade Decay Heat Removal (SGDHR) system capacity before the initiation of boiling. Sensitivity analysis of 500 MWe shows that the reactor goes to sub-critical and the transient power reduces to SGDHR system capacity before the boiling initiation. In the sensitivity analysis of 1000 MWe core, initiation of voiding and fuel melting occurs. But, with 80% core radial expansion reactivity feedback and nominal sodium expansion reactivity feedback, the reactor was maintained substantially sub-critical even beyond when net power crosses the SGDHR system capacity. From the study, it is concluded that if the sodium void reactivity is limited (4.6 $) then the inherent safety of 1000 MWe design is assured, even with 20% uncertainty on the sensitive parameters.     

Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor

The 500 MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam–Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1–0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than ∼1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using ‘CRAFT’ software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by β factor method.     

CARR应急堆芯冷却系统停堆冷却措施分析

停堆后的冷却问题是中国先进研究堆(CARR)重要的安全问题之一。CARR应急堆芯冷却系统是一套多功能、高度安全可靠的专设安全设施,它在反应堆正常运行时执行池水冷却功能;在正常停堆和事故停堆过程中执行应急堆芯冷却功能;还执行应急热阱选择、系统供电方式、回路阻力分析、阀门开关设置等方面的处理,使系统在两种功能的切换中不需要人为操作,依靠流量的自动匹配来满足正常运行和事故运行的要求。体现了CARR的安全性、先进性和经济性。本文以核安全法规和导则为前提,以满足系统功能为基础,首先介绍了CARR应急堆芯冷却系统的功能、主要参数和流程。根据CARR的实际情况,对应急堆芯冷却系统的停堆冷却措施和典型事故进行了分析,论证了该系统是如何在正常停堆和事故停堆状态下实现非能动堆芯冷却的。     

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.     

Hydraulic design of sparger arrangement for long guide tubes of a fast breeder reactor

In liquid metal cooled fast reactors, the core is submerged in sodium pool by ∼5 m below sodium free surface. This necessitates the control and shutdown of reactor be achieved by long overhanging mechanisms housed inside a control plug. These mechanisms are protected by porous guide tubes with a sparger type arrangement for the sodium flow through them. Comprehensive knowledge of flow distribution of sodium through these guide tubes is essential to assess the risks of flow induced vibration of thin thermowell tubes that pass close to these shroud tubes and entrainment of cover gas due to high free surface velocities. Three dimensional hydraulic analysis of single isolated shroud tube and integrated assembly of shroud tubes have been carried out using CFD tools to acquire this knowledge. The predictions of the CFD models have been validated against experimental predictions. These studies have provided important information regarding critical design parameters. Size of holes in the shroud tube, location of holes in the control plug shell and arrangement for breaking sodium jets emanating from shroud tubes have been optimized to reduce free surface velocity.     

Long operation life reactor for lunar surface power

The Pellet Bed Reactor (PeBR) with an operational life of 66 full-power years is developed for lunar surface power. It has Inconel X750 structure and vessel and would be launched unfueled then loaded with spherical fuel pellets (∼1.0 cm dia.) on the lunar surface after being placed below grade and surrounded with regolith. The pellets, comprised of ZrC-coated UC particles (∼850 μm in dia.) dispersed in ZrC matrix, are delivered to the lunar surface in subcritical canisters. The canisters are designed to remain sufficiently subcritical during launch and when submerged in wet sand and flooded with seawater in the unlikely event of a launch abort accident. The PeBR power system nominally generates ∼100 kW_e at a thermal efficiency of ∼21% and a reactor exit temperature of 910 K. It employs three separate closed Brayton cycle (CBC) loops each with a turbo-machine unit for energy conversion and two water heat pipes radiator panels for heat rejection. The reactor coolant and CBC working fluid is He-Xe binary gas mixture (40 g/mol). Estimates of the hot-clean excess reactivity and the full-power operation life are obtained using neutronics and fuel depletion analyses. In addition, estimates of the total radioactivity in post-operation PeBR, while being stored below grade on the lunar surface, are determined for up to 1000 years.     

Analyzing the alternative shutdown cooling behaviors for Chinshan Nuclear Power Plant using CFD simulation

The Chinshan Nuclear Power Plant (CSNPP) is a GE-designed BWR4 plant, having two identical units with rated core thermal power of 1804 MWt each unit. Several alternative shutdown cooling methods driven by natural or mixed convection has been proposed by the plant for studying the core cooling capability when the Residual Heat Removal (RHR) systems are not available or the refueling tasks, such as the In Vessel Visual Inspection (IVVI) work etc., is being proceeded. One of the examples is to connect a pipe from the outlet of the new spent fuel heat exchanger to the reactor cavity. The design of the alternatives shall ensure that the maximum core fluid temperature is limited below the boiling temperature of water. In this study, a Computational Fluid Dynamics (CFD) model is developed to analyze the natural convection phenomena during the shutdown period. Through a series of assumption, modeling and meshing processes, a calculation domain with approximate four million meshes including the RPV, reactor cavity and spent fuel pool, have been solved in this study. The analysis results showed that the passive alternative shutdown cooling system could provide sufficient heat removal capability to maintain the maximum core fluid temperature below boiling temperature. The results also indicated that the alternative shutdown cooling system could be served as an appropriate solution for CSNPP when the RHR is inoperable.     

Perspective channel-type reactor with enhanced safety

Following the search for new design solutions to develop within the framework of channel trends the reactor with enhanced safety the Research and Development Institute of Power Engineering has developed the design of the multiloop boiling water reactor (MKER). The MKER enhanced safety is attained when involving the inherent safety features, passive safety systems as well as the accident consequences confinement devices. The design realizes several advantages which are typical of the channel-type reactors, namely: the design desintegration simplifying the manufacture, control, equipment delivery and decreasing, versus the pressure vessel reactors, the accident effect if it proceeds in an explosive manner; small operating reactivity margin and fuel burnup increased due to continuous refuelling; fuel cycle flexibility allowing comparatively easily to adopt the reactor to the conjuncture of the country fuel balance; multiloop circuit of the main coolant which reduces the degree and effect of the accidents connected with the equipment and pipings rupture; monitoring of the channels and fuel assemblies leak-tightness.     

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.