Multiple Failure Accident Sequences for SEAFP Reactor

This paper summarizes several investigations on the identification of possible multiple failure accidents relevant in terms of consequences for the SEAFP reactor. Particularly, on those sequences of events that could induce a risk of radioactive materials bypass through the SEAFP confinement barriers. The analyses here reported are related to the Heat Transfer Systems of both reactor models 1 and 2. The work is carried out within the Safety and Environmental Assessment of Fusion Power—Long Term Programme (SEAL) 95/96. A set of specific initiating events (IEs) has been individuated, on the basis of the previous studies performed in the frame of the first SEAFP program. Basing on pre-existing analyses, each accident initiator has been discussed and several sequences have been described depending on the additional failures which could follow the initiator.     

EBT reactor magnetics and particle confinement

Optimization of the vacuum magnetic field of an ELMO Bumpy Torus (EBT) reactor is investigated. Several methods of improving reactor volume utilization and single particle confinement are analyzed. These include the use of (a) a large number of sectors and/or a large mirror ratio, (b) high field Nb₃Sn or Nb₃Sn/NbTi hybrid mirror coils, (c) split-wedge mirror coils, (d) axis-encircling aspect ratio enhancement (ARE) coils, and (e) recently developed field symmetrizing (SYM) coils. Of these, particle drift orbit and three-dimensional tensor pressure equilibrium calculations show that the use of SYM coils in conjunction with high field mirror magnets offers the most promise of good plasma performance in reactors that are smaller (by up to 50%) than previous reference designs that did not employ supplementary coils. Aspect ratio enhancement coils also offer an attractive alternative for improved confinement, but they do not have many of the advantages of SYM coils, particularly for reactor applications. Split-wedge mirror coils improve volume utilization and trapped particle confinement, but they do not enhance the confinement of transitional and passing particles. High field magnets improve confinement by permitting a larger mirror ratio and a larger plasma radius by virtue of their smaller cross-sectional area and higher current density. The relative merits of each magnetics configuration are discussed, including the effects on single particle confinement, reactor volume utilization, materials requirements, engineering design considerations, and reactor assembly, maintenance, and accessibility.     

Design and test of hydraulic vacuum breaker check valve for simplified boiling water reactor

In the current design of the simplified boiling water reactor, the vacuum breaker check valve is an important safety component. The vacuum breaker check valve is the only key safety components which is not passive in nature. Failure of this mechanical valve drastically reduces the passive containment cooling system cooling capability and hence containment pressure may exceed the design pressure. To eliminate this problem novel vacuum breaker check valve was developed to replace the mechanical valve. This new design is based on a passive hydraulic head, which is fail-safe and is truly passive in operation. Moreover this new design needs only one additional tank and one set of piping each to the wetwell and drywell. This system is simple in design and hence is easy to maintain and to qualify for operation. The passive vacuum breaker check valve performance was first evaluated using RELAP5. Then the passive vacuum breaker check valve was constructed and implemented in the PUMA integral test facility. Its performance was studied in a large break loss of coolant accident simulation test performed in PUMA facility.     

Fusion Specific Features in ITER Accident Analysis

Fusion specific features like inherent plasma shutdown, low decay heat densities, cryogenic temperatures, and limited source terms were considered during the safety design process of ITER. Uncertainties in plasma disruptions motivates a robust design to cope with multiple failures of in-vessel cooling piping. A vacuum vessel pressure suppression system mitigates pressure transients and effectively captures mobilized radioactivity. In case of pump trips or ex-vessel coolant losses in the divertor the plasma needs to be actively terminated in a few seconds. Failure to do so might damage the divertor but radiological consequences will be minor due to the intact first confinement barrier. Tritium plant inventories are protected by several layers of confinement. Uncontrolled release of magnet energy will be prevented by design. Postulated damage from magnets to confinement barriers causes fluid ingress (air, water, helium) into the cryostat. The cold environment limits pressurization. Most tritium and dust is captured by condensation.     

Integral analysis of large LOCA beyond design basis accident scenario in RBMK-1500

The integral analysis of severe accident scenario for RBMK-1500 was performed using combined approach with RELAP5, RELAP/SCDAPSIM, ASTEC and COCOSYS codes. The performed analysis covered response of the reactor core, the reactor cooling system and the confinement. There were performed several analyses: the first analysis assumed that operators take no action or their actions are not successful to provide the coolant injection to the reactor core; the other analyses were performed to investigate the accident management measures to restore the core cooling at different temperatures of the reactor core. The results of performed analyses showed that the operators have ∼5 h before the ruptures of fuel claddings occur and ∼8 h before the onset of exothermic steam-zirconium reaction. The coolant injection to the reactor core should be restored as soon as possible in order to prevent high hydrogen concentrations in the confinement and significant release of the fission products to the environment.     

The primary loop confinement and pressure boundary system of the HTR-10

The design philosophy and requirements of the HTR-10 reactor building and the primary loop confinement are introduced in this paper. Also introduced are the design, fabrication and the installation of the HTR-10 primary loop pressure boundary system. The primary loop confinement comprises the sealed cavities of the reinforced concrete structure. The main components and the connected gas systems of the primary loop pressure boundary system are contained in the confinement. Under normal operating condition, the inside pressure of the confinement is kept at negative pressure to ensure the sealing function of the confinement. There is a rupture disk of overpressure protection in the confinement wall. After a depressurization accident the pressure of the confinement increases and the rupture disk will break. The air of the confinement is discharged directly to the atmosphere through the accident discharge chimney which is connected to the rupture disk without filter. The main components of the primary loop pressure boundary system consist of the reactor pressure vessel, the steam generator pressure vessel and the hot gas duct vessel. All the above main components are installed in the reactor cavity and the steam generator cavity. They are all nuclear safety class 1 components, whose materials production, design, fabrication, and tests are carried out according to ASME Section III and relevant Chinese nuclear codes.     

Design of the exhaust system in the confinement of the HTR-10

The modular high temperature gas-cooled reactor has a vented confinement instead of a gastight pressurized containment due to its passive safety features. The safety class negative pressure exhaust system is used in the heating, ventilation and air conditioning system to fulfill all kinds of safety-related functions at the normal operation and during accidents. This paper introduces and reviews the design of safety class negative pressure exhaust systems of the 10 MW high temperature gas-cooled reactor-test module.     

Cooling System Design Options for a Fusion Reactor

The objective of a fusion power reactor is to produce electricity safely and reliably. Accordingly, the design, objective of the heat transport system is to optimize power production, safety, and reliability. Such an optimization process, however, is constrained by many factors, including, among others: public safety, worker safety, steam cycle efficiency, reliability, and cost. As these factors impose conflicting requirements, there is a need to find an optimum design solution, i.e., one that satisfies all requirements, but not necessarily each requirement optimally. The SEAFP reactor study developed helium-cooled and water-cooled models for assessment purposes. Among other things, the current study demonstrates that neither model offers an optimum solution. Helium cooling offers a high steam cycle efficiency but poor reliability for the cooling of high heat flux components (divertor and first wall). Alternatively, water cooling offers a low steam cycle efficiency, but reasonable reliability for the cooling of such components. It is concluded that an optimum solution includes helium cooling of low heat flux components and water cooling of high heat flux components. Relative to the SEAFP helium model, this hybrid system enhances safety and reliability, while retaining the high steam cycle efficiency of that model.     

模块式小堆超压风险及设计优化研究

模块式小堆采用带直流蒸汽发生器（OTSG）的一体化堆芯设计。OTSG具有传热面积大、设备体积小、蒸汽品质高的优点,然而因其二次侧水装量小、热惯性差,当反应堆发生二次侧排热减少时,反应堆冷却剂系统（RCS）可能存在超压风险。紧凑的一体化布置使得堆芯应对冷却剂受热膨胀的能力减弱,进一步增大RCS超压风险。本文采用RELAP5程序对模块式小堆的超压风险进行了研究。研究结果表明,模块式小堆在二次侧排热减少事故中会出现RCS超压现象,其中汽轮机事故停机导致的超压后果最为严重。波动管的流通面积对于RCS压力有着显著影响,合理地设计波动管流通面积可缓解RCS超压。     

Overview of the SEAFP and SEAL Studies

The SEAFP (Safety and Environmental Assessment of Fusion Power) and SEAL (Safety and Environmental Assessment of fusion power, Long-term) programs form part of the ongoing effort in the European Fusion Programme to consider the safety and environmental aspects of fusion power. SEAFP was undertaken in the period 1992–1994. The assessment started with the development of two conceptual power plant designs, each of 3000 MW of fusion power, termed Model 1 and Model 2. Model 1 used vanadium alloy, helium cooling, and lithium oxide for tritium generation. Model 2 used a reduced-activation martensitic steel, water cooling, and a lithium–lead alloy for tritium generation. Both Models were designed for passive safety. The SEAFP analyses included detailed consideration of effluents, occupational doses, accidents (concentrating on the worst possible accidents), and waste management. The key results are summarized in this paper. SEAL was launched in 1995, with the aims of broadening the scope of SEAFP, and of elaborating selected aspects of SEAFP in more detail. The SEAL analyses include studies which extend the results of SEAFP to a wider class of blanket designs and material choices, improved assessments of the quantities of activated materials which may be exempted from regulatory control or recycled, improved modeling of occupational doses, and work in many areas to improve relevant data, modeling and analyses, or consider design improvements. Much of this work is ongoing, but key results from completed work are summarized in this paper.     

The design and experimental validation of an emergency core cooling system for a pool type research reactor

This paper presents the design of the Emergency Core Cooling System (ECCS) for the IEA-R1m pool type research reactor. This system with passive features, uses sprays installed above the core. The experimental program performed to define system parameters and to demonstrate to the licensing authorities, that the fuel elements limiting temperature is not exceeded, is also presented. Flow distribution experiments using a core mock-up in full-scale were performed to define the spray header geometry and spray nozzles specifications as well as the system total flow rate. Another set of experiments using electrically heated plates simulating heat fluxes corresponding to the decay heat curve after full power operation at 5 MW was conducted to measure the temperature distribution at the most critical position. The observed water flow pattern through the plates has a very peculiar behavior resulting in a temperature distribution which was modelled by a 2D energy equation numerical solution. In all tested conditions the measured temperatures were shown to be below the limiting value.     

RAMI Analysis Program Design and Research for CFETR (Chinese Fusion Engineering Testing Reactor) Tokamak Machine

Chinese Fusion Engineering Testing Reactor (CFETR) is a test reactor which shall be constructed by National Integration Design Group for Magnetic Confinement Fusion Reactor of China with an ambitious scientific and technological goal. The reactor has the equivalent scale compared with ITER, but has the complementary function to ITER. CFETR is a demonstration of long pulse or steady-state operation with duty cycle time not less than 0.3–0.5 and the full cycle of tritium self-sustained with TBR not less than 1.2. At the same time it will be exploring options for DEMO blanket and divertor with an easy changeable core by remote handling way. To be able to reach its scientific and technological objectives, as one of technical risks control methods, RAMI analysis need to be done during the hold lifetime of CFETR, from conception design to decommissioning. Base on stating of CFETR lifetime and preliminary operational programme, the RAMI analysis program and process are designed and discussed, it consists of five major steps: (1) functional analysis are performed, (2) calculating reliability block diagrams, (3) analyzing failure mode, effects and criticality analysis, (4) risk mitigation actions are taken to ensure every system is compatibility with RAMI objectives, (5) All the RAMI analysis are integrated as the final RAMI analysis reports to be reviewed in the system final design review. Along with the elements of the analysis the vacuum vessel (VV) system was performed to provide as examples, detailed showing how the CFETR RAMI analysis is carried out. CFETR RAMI analysis guidelines were designed and established, after constantly revised and improved these analysis criteria and programs will become the basis standards for CFETR RAMI analysis. Preliminary RAMI analysis of CFETR VV system was obtained, which will be updated with the VV system design progresses.     

Analysis of the reference accidental sequence for safety assessment of LLCB TBM system

Safety analysis of the reference accidental sequence has been carried out for Lead Lithium cooled Ceramic Breeder (LLCB) Test Blanket Module (TBM) system; India's prototype of DEMO blanket concept for testing in International Thermonuclear Experimental Reactor (ITER). The accidental event analyzed starts with a Postulated Initiating Event (PIE) of ex-vessel loss of first wall helium coolant due to guillotine rupture of coolant pipe with simultaneous assumed failure of plasma shutdown system. Three different variants of the sequences analyzed include simultaneous additional failures of TBM and ITER first wall, failure of TBM box resulting in to spilling of lead lithium liquid metal in to vacuum vessel and reactor trip on Loss of Coolant Accident (LOCA) signal from TBM system. The analysis address specific reactor safety concerns, such as pressurization of confinement buildings, vacuum vessel pressurization, release of activated products and tritium during these accidental events and hydrogen production from chemical reactions between lead–lithium liquid metal and beryllium with water. An in-house customized computer code is developed and through these deterministic safety analyses the prescribed safety limits are shown to be well within limits for Indian LLCB-TBM design and it also meets overall safety goal for ITER. This paper reports transient analysis results of the safety assessment.     

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。      

10 MW高温气冷实验堆事故分析的结果与对策

1 0MW高温气冷实验堆 (HTR 1 0 )的事故分析表明 ,在设计基准事故和严重事故条件下 ,HTR 1 0的堆芯燃料元件的最高温度和反应堆冷却剂系统的压力都低于规定的安全限值 ,燃料元件和冷却剂系统压力边界都能保持其完整性 ,不会造成裂变产物大量向外释放。根据事故分析结果并参照国外高温气冷堆安全运行的管理实践经验 ,针对HTR 1 0所提出的一系列事故对策有效地保证了HTR 1 0在较高的安全水平上进行设计、建造、运行及管理等 ,能够确保HTR 1 0、人员、社会以及环境的安全     

The design features of the HTR-10

The 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is a modular pebble bed type reactor. This paper briefly introduces the main design features and safety concept of the HTR-10. The design features of the pebble bed reactor core, the pressure boundary of the primary circuit, the decay heat removal system and the two independent reactor shutdown systems and the barrier of confinement are described in this paper.     

RAPID Operator-Free Fast Reactor Concept without Any Control Rods Reactor Concept and Plant Dynamics Analyses

The 1,000kWe metal fueled sodium-cooled fast reactor concept “RAPID” to achieve highly automated reactor operation has been demonstrated. RAPID (Refueling by All Pins Integrated Design) is designed for a terrestrial power system which enables quick and simplified refueling. It is one of the successors of the RAPID-L, the operator-free fast reactor concept designed for lunar base power system. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 14,000 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years.

Unique challenges in reactivity control systems design have been addressed in the RAPID concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt ⁶Li as a liquid poison instead of B₄C rods. In combination with LEMs, LIMs and LRMs, RAPID can be operated without an operator. In this paper, the RAPID reactor concept and its transient characteristics are presented.     

Core Safety Characteristics of High Flux Safe Reactor (HFSR)

Abstract

Transient analyses are performed for graphite moderated helium-cooled high flux reactor to obtain the high flux safe reactor design. In order to promote the safety of the high flux reactor, the present design adopts the pebble bed reactor and its fuel technology. In the transient analyses, among the postulated off-normal events and accidents, the reactivity accident followed by a loss of helium forced circulation with system depressurization is found to be the severest potential event which may threaten the reactor safety from fission products release point of view. Several neutronic and thermal-hydraulic design parameters are indicated and exploited to promote the reactor safety. Neutronic and core thermal-hydralic models are proposed and used to simulate the reactor responses to the off-normal events and accidents. As the results of the transient analyses and accident simulation, safe and optimal design parameters are obtained which provide high thermal neutron flux with a desirable spectrum and large usable volume constrained by safety limitations.     

A quantitative methodology for reactor vessel pressurized thermal shock decision making

The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). Previous reactor vessel integrity concerns have led to changes in vessel and plant system design and to operating procedures, and increased attention to the PTS issue is causing consideration of further modifications. Events such as excess feedwater, loss of normal feedwater, and steam generator tube rupture have led to significant primary system cooldowns. Each of these cooldown transients occurred concurrently with a relatively high primary system pressure. Consideration of these and other postulated cooldown events has drawn attention to the impact of operator action and control system effects on reactor vessel PTS.A methodology, which couples event sequence analysis with probabilistic fracture mechanics analyses, was developed to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation.Once the specific event sequences of concern are identified, detailed thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. This paper addresses key aspects of the thermal-hydraulic and fracture mechanics analyses of the reactor vessel. The effects of incomplete mixing of safety injection flow in the primary cold leg and vessel downcomer and the application of warm prestressing are emphasized. The results of these analyses are being used to define further modifications in vessel and plant system design and to operating procedures.Previous design considerations that have evolved as a result of reactor vessel integrity evaluations are mentioned. These include the development of realistic design analysis tools and selection of plant system modifications. Modifications that are being developed or are under consideration are also mentioned. These include vessel fluence reductions, additional modifications to operating procedures, increased use of probabilistic event sequence and fracture mechanics analysis methods, enhanced material fracture toughness, and reductions in the severity or frequency of occurrence of dominant reactor vessel PTS transients.     

Stress and Thermal Analysis of the In-Vessel RMP Coils in HL-2M

A set of in-vessel resonant magnetic perturbation(RMP) coils for MHD instability suppression is proposed for the design of a HL-2M tokamak.Each coil is to be fed with a current of up to 5 kA,operated in a frequency range from DC to about 1 kHz.Stainless steel(SS) jacketed mineral insulated cables are proposed for the conductor of the coils.In-vessel coils must withstand large electromagnetic(EM) and thermal loads.The support,insulation and vacuum sealing in a very limited space are crucial issues for engineering design.Hence finite element calculations are performed to verify the design,optimize the support by minimizing stress caused by EM forces on the coil conductors and work out the temperature rise occurring on the coil in diferent working conditions,the corresponding thermal stress caused by the thermal expansion of materials is evaluated to be allowable.The techniques to develop the in-vessel RMP coils,such as support,insulation and cooling,are discussed.