大功率先进压水堆压力容器外部冷却能力研究

目前压力容器外部冷却（ERVC）作为严重事故管理策略中压力容器内熔融物滞留（IVR）的一部分已得到了广泛应用。本文采用RELAP5系统安全分析程序定性研究一些流动参数和边界条件（如进出口面积、冷却水的入口温度、下封头处的加热功率、下封头处流道的间隙尺寸及注水高度等）对大功率先进压水堆压力容器外部冷却的自然循环能力产生的效应,它为结构的设计和系统的瞬态响应行为提供了一定的分析依据。     

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation.     

An application of the new way to prevent core melting in pressure tube reactors (CANDU type)

The pressure tube reactors, especially CANDU type, have a calandria low pressure vessel (near to atmospheric pressure) immersed into a concrete vault filled with water. The accident analysis done by ELFIN-HTCELL code for the channel heat up and by fluid flow PHOENICS code as applied for moderator cooling system efficacy, showed that even the moderator cooling system operates, in some transients sequences where the normal heat sinks are lost, and the top core pressure tubes can reach burst conditions, which means that the fission product secondary retaining barrier gets destroyed, and yet the core can be cooled by water admission through the ruptured tubes from the emergency core cooling system (ECC), if it is available. Otherwise, if in many accident sequences the moderator cooling system remains the ultimate heat sink for the core fuel, and it is not available even from the accident start, a core melt appears. Taking into account the “natural” advantage offered by the presence of both pools in calandria and in the vault, separated by the calandria vessel, the introduction of density locks between them could be a safety passive design solution. When the temperature of moderator water gets higher the density lock cold-hot interface loss stability and thus the density locks get “open” fully permitting the admission of the cool water from the vault pool in calandria. Therefore, by natural circulation the decay heat is transferred via an air-cooling tower, and no mechanical moving parts are needed to open this circuit. Also, if the vault water is borated, it can be used to stop the nuclear reaction when the normal shutdown systems are not available and a positive reactivity coefficient appears, e.g. large loss of coolant accident (LOCA).     

Pressure transients in NET building compartments under LOCA conditions

The NET cooling systems for in-vessel components and vessel are generally based on low pressure and low temperature water. However, the cooling loops for the breeder blanket are intended to operate at a water temperature of about 250°C. A pipe break in a loop with such data would pressurize the compartment where the break takes place. Therefore, as a basis for proper compartment design, it is important to analyze possible pressure increases following pipe breaks. It may also be necessary to introduce equipment for pressure relief or pressure suppression. The objective of the parameter study presented is to determine the relationship between allowed maximum containment pressure following postulated large pipe break in breeding blanket loop and required containment volume. Parameters varied are: blanket loop temperature and pressure (within the range of burn and baking), and pressure suppression system inclusion/exclusion. The analysis has been performed by means of the Swedish containment code COPTA. The results of the analysis are summarized in a plot showing the influence of the varied parameters on required containment volume. In addition, the results presented include required vent areas, heat sink capacities, etc.     

Heat Removal Performance Evaluation of Several Passive Containment Cooling Systems during Loss of Coolant Accident

As a passive containment cooling system (PCCS), which is adopted in simplified BWRs, several concepts, differing in cooling location and method, such as the suppression chamber water wall, the drywell water wall, the isolation condenser (I/C) and the drywell cooler, have been considered. This paper summarizes the characteristics of each PCCS concept, and the analysis results of the performance for several PCCSs during a main steam line break LOCA for a reference simplified BWR plant, obtained by the newly developed containment thermalhydraulic response analysis code TOSPAC.

The performance comparison suggests that I/C and drywell cooler have good heat removal capability with regard to the smallest heat transfer area among PCCS concepts evaluated in the present analysis. I/C removes decay heat efficiently, since it absorbs steam directly from the reactor pressure vessel, which is the hottest portion inside the containment. The suppression chamber water wall is ineffective, mainly due to high non-condensable gas partial pressure in the suppression chamber, and low suppression pool temperature.

Calculations of other pipe breaks were also implemented for the reference plant adopting I/C as PCCS. The results show the effectiveness of the I/C cooling over a wide range of break spectra.     

Development of Severe Accident Analysis Code SAMPSON in IMPACT Project

IMPACT is the name of a program and of specific simulation software, which will perform full-scope and detailed calculations of various phenomena in a nuclear power plant for a wide range of event scenarios. The four years of the IMPACT project Phase 1 have been completed, and each analysis module of the prototype version of the severe accident analysis code SAMPSON has been developed and verified by comparison with separate-effect test data. Verification of the integrated code with combinations of up to 11 analysis modules has been conducted, with the Analysis Control Module, to demonstrate the code capability and integrity. A 10-inch cold leg failure Loss of Coolant Accident in the Surry Plant was the assumed initiating event. The system analysis was divided into two cases; one was an in-vessel retention analysis when gap cooling was effective, the other was an analysis of phenomena when the event was extended to ex-vessel due to the reactor pressure vessel failure when gap cooling was not sufficient. Using the Analysis Control Module to select and execute adequate combinations of the various analysis modules dynamically according to the progression of plant phenomena and to control parallel processing, the goal of integrated calculations by SAMPSON with multiple analysis modules executing in parallel was achieved.     

Development of TSACO and application to Chinese HCCB TBM cooling system

In this study, a thermal-hydraulic and safety analysis code (TSACO) for helium cooling system has been developed using Fortran 90 language, and the simulation has been performed for the cooling system of the Chinese helium cooled ceramic breeder test blanket module (CH HCCB TBM). The semi-implicit finite difference technique was adopted for the solution of the dynamic behavior of helium cooling system. Furthermore, a detailed illustration of the numerical solution for heat structures and critical model was presented. The code was verified by the comparison of RELAP5 code with the same initial condition, boundary condition, heat transfer and flow friction models. The TBM inlet/outlet temperatures and pressure drop were obtained and the results simulated by TSACO were shown in good agreement with those by RELAP5. Thereafter, the design basis accident in-vessel loss of coolant accident (LOCA), was investigated for the CH HCCB TBM cooling system. The critical flow model was also verified by comparing with RELAP5 code. The results indicated that the TBM can be cooled down effectively. The vacuum vessel (VV) pressure and the mass of helium spilled into the VV maintained below the design limits with a large margin.     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

The HTTR reactor pressure vessel and its integrity against a PTS event

Before manufacturing the real steel to be used in the reactor pressure vessel (RPV) of the high temperature engineering test reactor (HTTR) the vessel manufacturer and materials supplier made a sample steel by the same procedure as for the real steel (2.25Cr-1 Mo) and conducted many tests to obtain material strength data for its base and weld metals. The test results showed that the sample steel satisfied the HTTR design requirements. Vessel cooling panels are set on the inner surface of the biological shielding concrete around the RPV, and are circulated with cooling water at 0.5 MPa and 40°C to cool the shielding concrete during normal operation of the reactor. By supposing that the cooling panel breakes and the water discharges to the RPV outer surface heated at 400°C, the stress distribution generated in the vessel wall by a pressurized thermal shock (PTS) event can be calculated using a finite element method code. This paper describes some of the results obtained from the material testing of the sample steel and the estimated result using the scheme developed for a light water reactor pressure vessel, to clarify the integrity of the HTTR-RPV under a PTS event.     

Simulation of melt jet breakup and debris bed formation in water pools with IKEJET/IKEMIX

The objective of the development of the code system KESS is simulating the processes of core melting, relocation of core material to the lower head of the reactor pressure vessel (RPV) and its further heatup, modelling of fission product release and coolability of the core material. In the scope of the code development, IKEJET and IKEMIX were designed as key models for the breakup of a molten jet falling into a water pool, cooling of fragments and the formation of particulate debris beds. Calculations were performed with these codes, simulating FARO corium quenching experiments at saturated (L-28) and subcooled (L-31) conditions, as well as PREMIX experiments, e.g. PM-16. With the assumption of a reduced interfacial friction between water and steam as compared to usually applied laws, the melt breakup, energy release from the melt and pressurisation of the vessel observed in the experiments are reproduced with a reasonable accuracy. The model is further applied to reactor conditions, calculating the relocation of a mass of corium of 30 t into the lower plenum, its fragmentation and the formation of a particle bed.     

Fracture mechanical evaluation of an in-vessel melt retention scenario

This paper presents methods to compute J-integral values for cracks in two- and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.     

Assessment of reactor vessel integrity (ARVI)

The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants.The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels, (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.     

Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40?mm inner diameter coolant pipes providing water for the divertor targets during the “baking” regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. “Baking” regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal–hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1?s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.     

Three-dimensional analysis of flow characteristics on the reactor vessel downcomer during the late reflood phase of a postulated LBLOCA

A three-dimensional CFD analysis has been performed on the flow characteristics in the reactor vessel downcomer during the late reflood phase of a postulated large-break loss-of-coolant accident (LBLOCA), in order to validate the modified linear scaling methodology that was applied in the MIDAS test facility of Korea Atomic Energy Research Institute. The vertical and circumferential velocity similarities are numerically tested for the 1/1 and 1/5 linear scale models for the APR1400 reactor vessel downcomer. The effects of scale on flow patterns, pressure and velocity distributions, and the impinging jet behavior are analyzed with the FLUENT code. In addition, a simplified half cylinder model with a single emergency core cooling (ECC) nozzle is numerically tested to investigate the scale effect on the spreading width and break-up of ECC water film. The qualitative and quantitative results indicate that the 1/5 modified linear scale model of the reactor vessel downcomer would reasonably preserve the hydrodynamic similarity with APR1400.     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间,从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象,采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效,堆芯因冷却不足升温熔化导致压力容器失效,熔融物与混凝土发生反应（MCCI）,安全壳超压失效;冷管段大破口叠加再循环失效,安全壳内蒸汽不断聚集,发生超压失效。通过对两种事故工况的分析,证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。     

Safety analysis of multiple-failure of passive safety systems in SBWR-1200 SBLOCA

The design of the simplified boiling water reactor (SBWR-1200) is characterized by utilizing fully passive safety systems. The emergency core cooling is realized by the gravity driven core cooling system, and the decay heat removal is done by the passive containment cooling system and isolation condenser system. All of the systems have multiple units and could be partially failed. The objective of this paper is to analyze the system response under the multiple malfunctions of passive safety systems in the SBWR-1200.

The chosen accident scenario is a small break loss of coolant accident with one of three gravity driven core cooling system drain lines blocked and one of three passive containment cooling system condensers disabled. An integral test has been carried out in the PUMA facility for 16 h. The facility is designed for low pressure, long term cooling operation with the multiple safety related components; therefore, it has the flexibility to demonstrate the asymmetric or multiple-failure effects with the combination of disability of safety systems. The test initial conditions at 1 MPa (150 psi) are obtained from RELAP5/MOD3.2 code simulation for the SBWR-1200 with appropriate scaling considerations.

Comparisons have been first made between the multiple-failure test and a single-failure test preformed previously. It shows that the core has been covered with liquid coolant during all of accident transient even though there is an apparent coolant inventory reduction in the multiple-failure test. The decay heat removal has no significant difference because the remaining two passive containment cooling condensers increase their cooling capacities, and even the drywell pressure is slightly lower due to the cold water injection from the suppression pool. Comparisons have also been made between the scaled-up test data and the code simulation at the prototypic level. The prototypic simulation is done by RELAP5/MOD3.2. Agreements between the code simulation and the scaled-up test data confirm the code applicability and the facility scalability for this accident scenario.     

Modeling of a confinement bypass accident with CONSEN,a fast-running code for safety analyses in fusion reactors

The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code.     

针对REPEC加热实验的RELAP5程序模拟与分析

为研究压力容器外部流道的冷却能力及流动传热过程,在反应堆压力容器外部冷却(REPEC, Reactor Pressure vessel External Cooling)实验台架前期加热实验的基础上,采用RELAP5程序对实验工况进行模拟和对比。模拟结果与实验数据一致性较好。随加热热流、进出口面积的增加,系统内自然循环流量也增加;入口欠热度对自然循环流量的影响不是很明显;近饱和沸腾条件下,系统出现明显的两相不稳定流动。     

Basic concept of a near future BWR

This paper shows a basic concept of a near future boiling water reactor (BWR) aiming at evolutional safety and cost savings with minimum change from the current advanced BWR (ABWR). The plant output is uprated to 1500 MWe from 1356 MWe. This power uprate can bring about potential of 11% cost saving per MWe base. Safety improvement as a next generation large reactor is also achieved.

The advanced reinforced concrete containment vessel (ARCCV) is used for the containment vessel to improve safety for severe accidents. The peak pressure of the containment at severe accidents can be kept close to the design pressure. The advanced passive containment cooling system (APCS) is also provided and can accomplish no primary containment vessel (PCV) venting.

The advanced emergency core cooling system (AECCS) consists of four divisions in the front line. The advanced passive cooling system (APCS) is also provided. The combination of the four divisional emergency core cooling system (ECCS) and the passive safety system improves the plant performance in probabilistic safety assessment (PSA).

This plant concept is designed based on the heritage of the current ABWR. No more major research and development (R&D) are necessary. Therefore, construction and operation is possible in the early 2010s.     

A thermophysical heavy water properties package for thermal-hydraulic accident analysis codes

A thermodynamic and transport properties package for heavy water (D₂O) has been prepared. This package has been implemented in an advanced Nuclear Reactors Thermal-Hydraulic accident analysis code. Several practical problems are analyzed and a comparison between D₂O and H₂O as cooling agents is presented.