A theoretical study of cyclon-effect in PWR downcomer during a LOCA blowdown

The current strategy based on the use of so-called cross-junction, allows partial modeling during blowdown episodes driven by larger-scale (flow features such as helical profile) in downcomers nuclear reactors. However a subtle but significant effect may appear by the combined action of two factors: on the one hand high azimuthal flow, on the other hand the intrinsic curvature of downcomer, and additionally in presence of a two-phases (vapor–liquid) a cyclon-effect can manifest. The present paper is a theoretical analysis of a possible cyclon-effect during blowdown episodes that allows a qualitative estimate of the impact on the calculations.     

Experimental and post-test calculation results of the integral reflood test QUENCH-12 with a VVER-type bundle

The QUENCH-12 experiment was carried out to investigate the effects of VVER materials (niobium-bearing alloys) and bundle geometry on core reflood, in comparison with test QUENCH-06 using western PWR materials (Zircaloy-4) and bundle geometry. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1450 K, followed by a power ramp until a temperature of 2050 K was reached, then reflood with water at room temperature was initiated. The total hydrogen production was 58 g (QUENCH-06: 36 g), 24 g of which were released during reflood (QUENCH-06: 4 g). Reasons for the increased hydrogen production may be extensive damaging of the cladding surfaces due to the breakaway oxidation and local melt formation with subsequent melt oxidation. Post-test videoscope observations and metallographic investigations showed an influence of the breakaway oxidation with extensive spalling of oxide scales of rod claddings, shroud and auxiliary corner rods. The hydrogen content in the corner rods, withdrawn from the bundle during the test, reached more than 30 at% at the bundle elevations of 850 and 1100 mm. Post-test calculations were performed with local versions of SCDAP/RELAP5 following on from pre-test analyses with SCDAP/RELAP5 and SCDAPSIM.     

Behaviour of nonlinear supports on a PWR coolant system during a postulated LOCA: Part 1: Effect of modelling methods

A 4-loop Pressurised Water Reactor (PWR) primary coolant system has been analysed for the postulated Loss of Coolant Accident (LOCA) event in order to derive peak dynamic loads for qualifying the design of equipment supports and pipe whip restraints. Pipe whip restraints as well as pipe and equipment supports are nonlinear by nature because of the presence of gaps and the different directional stiffnesses arising from snubber, steelwork and geometric and material interaction at the concrete to steel embedment. The different structural idealisations for the supports and restraints have an influence on the dynamic response of the structure. In the first of the two part paper a range of idealisation models for the Steam Generator and Reactor Coolant Pump vertical columns ranging from elastic stiffnesses to bilinear stiffnesses with or without preload were examined. Due to both structural and loading complexity, the behaviour of these supports were analysed by the Nonlinear Modal Superposition Method [1] and [2]. Analyses with the different models that are within the capabilities of the WESTDYN [3] piping analysis code enabled variations in the peak dynamic loads at the supports, restraints and equipment nozzles and other parameters to be determined and the design acceptability of the hardware established.     

Experimental and calculation results of the integral reflood test QUENCH-15 with ZIRLO™ cladding tubes in comparison with results of previous QUENCH tests

The QUENCH-15 experiment investigated the effect of ZIRLO™¹ cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (standard Zircaloy-4), QUENCH-12 (VVER, E110), and QUENCH-14 (M5^®). The QUENCH-15 bundle cross-section corresponded to a Westinghouse PWR core design and consisted of 24 heated rods (internal tungsten heaters between 0 and 1024 mm axial elevation, cladding oxidised region between −470 and 1500 mm), six corner rods made of Zircaloy-4, two corner rods made of E110, and a Zirconium 702 shroud. The test was conducted in principle with the same protocol as QUENCH-06, -12 and -14, so that the effects of the change of cladding material and bundle geometry could be more easily observed. The test protocol involved pre-oxidation to a maximum of about 150 μm oxide thickness at a temperature of about 1473 K over a period of about 3000s. The power was then ramped at a rate of 0.25 W/s/rod to cause a temperature increase until the desired maximum bundle temperature of 2153 K was reached. The maximum oxide layer thickness observed was 380 μm. Then reflood with 1.3 g/s/rod water at room temperature was initiated. The electrical power was reduced to 175 W/rod during the reflood phase, approximating effective decay heat level. The post-test metallography of the bundle showed neither noticeable breakaway oxidation of the cladding nor melt release into space between rods. The average outer oxide layer thickness at hottest elevation of 950 mm was 620 μm (QUENCH-06: 630 μm). The molten cladding metal at hottest elevation was localised between the outer and inner oxide layers. The thickness of inner oxide layer reaches 20% of that of the outer oxide layer. The measured hydrogen release during the QUENCH-15 test was 41 g in the pre-oxidation and transient phases and 7 g in the quench phase which are comparable with those in QUENCH-06, i.e. 32 g and 4 g, respectively. Post-test calculations were performed using a version of SCDAP/RELAP5/MOD3.2. The calculation results support the heuristic observation that there was no major difference between the influence of Zircaloy-4, M5^® or ZIRLO™ for the beyond-design basis accident present conditions here studied.     

3-D numerical investigation of water/CuO nanofluid critical heat flux phenomenon in a PWR core channel during LOCA

Forced convection boiling and critical heat flux have been under considerable attention in variety of areas due to high heat removal capacity. However, once the heat flux exceeds a certain high level (CHF), the heated surface can no longer support continuous liquid contact, associated with substantial reduction in the heat transfer efficiency. One way to increase the level of the CHF is to add certain nanoparticles to the base fluid. The present paper investigates the effects of the addition of copper oxide nanoparticles on CHF phenomenon within the general-purpose computational fluid dynamics (CFD). The governing equations solved are generalized phase continuity, momentum and energy equations. Wall boiling phenomena are modeled using the baseline mechanistic nucleate boiling model developed in Rensselaer Polytechnic Institute (RPI). To simulate the critical heat flux phenomenon, the RPI model is extended to the departure from nucleate boiling (DNB) by partitioning wall heat flux to both liquid and vapor phases considering the existence of thin liquid wall film. It was shown that the presence of copper oxide nanoparticles in the base fluid, delays the dryout phenomenon dramatically and in specific concentration, CHF threshold would be enhanced, therefore, raising the upper limit of CHF could allow for higher safety margins.     

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.     

A study of fuel behavior in PWR design basis accident: an analysis of results from the PHEBUS and EDGAR experiments

As part of the French PWR safety study programme, fuel behavior during a design basis accident has been investigated in three parallel directions:

• - separate effect tests in the EDGAR apparatus for developing and validating cladding deformation models,

• - integral tests in PHEBUS for verifying codes,

• - development of fuel behaviour codes for plant calculation after assessment against experimental results. After describing the objectives and content of each of these programmes, the main findings are highlighted and discussed.

基于RELAP5的LOCA喷放阶段下降段内CCFL特性研究

反应堆失水事故（LOCA）后下降段通道内形成的两相逆流状态极有可能引发汽-液逆向流动限制（CCFL）,不利于应急冷却水顺利进入堆芯,极大影响了核反应堆系统的安全性能。本研究基于RELAP5程序采用Wallis溢流关系式对UPFT实验装置进行建模并计算LOCA喷放阶段的下降段注水行为;通过对比下腔室蓄水量、下降段内压力及破口处蒸汽流量瞬态变化以验证模型的有效性,并对下降段通道内汽相速度场、液相体积分数分布特性进行分析。结果表明,由于下降段通道结构的三维特征引起的流动不均匀性影响了汽-液CCFL特性,随着蒸汽流量增大,在破口环路与下降段连接区域的压力梯度与向上流速度梯度越大,较少节点的划分方法很难真实反映下降段通道局部区域内汽-液溢流关系;在靠近破口的环路内注入的冷却水更难到达下腔室,而在远离破口环路的冷却水容易进入到下腔室;过热的蒸汽在流动过程中被冷却水冷却发生凝结现象,导致出口蒸汽流量小于进口蒸汽流量,且随着进口蒸汽流量的增大,凝结效应则随之减小。本研究所建立的模型与方法能够适用于LOCA喷放阶段下降段通道内的汽-液CCFL预测。      

The effect of oxidation-and-quenching process during a LOCA on the behavior of the oxidation and embrittlement of Zircaloy-4 cladding under reheating transients

Two-sided oxidation tests, ring compression tests and semi-integral quench tests on Zircaloy-4 cladding specimens were conducted under temperature transient conditions simulating a post-quench reheat transient in order to evaluate the effect of high-temperature oxidation and quenching during a loss-of-coolant accident (LOCA) on the behavior of the oxidation and embrittlement of the cladding under a loss of long-term core-cooling condition. Test specimens prepared from non-irradiated Zircaloy-4 cladding tube were oxidized at a temperature between 1173 and 1473 K in steam flow and quenched by soaking the specimen in room temperature water. Re-heating tests were performed on the specimens in steam flow at a temperature between 1173 and 1473 K. The suppression of oxide layer growth and weight gain was observed under certain reheating-after-quenching conditions. Nevertheless, it seemed that the temperature transients including quenching-and-reheating process did not significantly affect the embrittlement of cladding. It was found that the embrittlement behavior of cladding during the temperature transients including quenching-and-reheating process could be dealt with on the basis of the Equivalent Cladding Reacted (ECR) based on the Baker–Just correlation.     

Factor analysis on hazards for safety assessment in decommissioning workplace of nuclear facilities using a semantic differential method

The decommissioning of nuclear facilities must be accomplished according to its structural conditions and radiological characteristics. An effective risk analysis requires basic knowledge about possible risks, characteristics of potential hazards, and comprehensive understanding of the associated cause-effect relationships within a decommissioning for nuclear facilities. The hazards associated with a decommissioning plan are important not only because they may be a direct cause of harm to workers but also because their occurrence may, indirectly, result in increased radiological and non-radiological hazards. Workers need to be protected by eliminating or reducing the radiological and non-radiological hazards that may arise during routine decommissioning activities as well as during accidents. Therefore, to prepare the safety assessment for decommissioning of nuclear facilities, the radiological and non-radiological hazards should be systematically identified and classified. With a semantic differential method of screening factor and risk perception factor, the radiological and non-radiological hazards are screened and identified.     

Numerical analysis of steam–air behavior in a pressurizer during reflux cooling

During reflux cooling, proper evaluation of behavior of accumulated non-condensable gases in the steam generator (SG) U-tubes is important to predict the performance of the reflux cooling. Non-condensable gases are present in the pressurizer and the possibility of migration of air in the pressurizer to the SG U-tubes is not well known. Steam and air behavior in the pressurizer during reflux cooling was, therefore, analyzed numerically using FLUENT 6.3.26 and the possibility of migration of air to the hot leg was investigated. For the analysis, the pressurizer of the ROSA-IV/LSTF experiment was employed as a calculation domain, since experimental data about the loss of the residual heat removal event during mid-loop operation are available. Two stages were assumed. (1) Phase 1: latent heat accumulated in the wall of the pressurizer and was eventually released to the outside. (2) Phase 2: the wall was heated up to the saturated steam temperature, and only heat loss to the outside occurred. The prediction indicated that in Phase 1, the air did not migrate to the surge line in either laminar or turbulent flow calculations, while in Phase 2 the air migrated into the hot leg only in the laminar flow calculation. Judging from a previous experiment on an axisymmetric free jet, the flow pattern in the pressurizer seems to be turbulent. In addition, a comparison of the analytical results of the fluid temperatures near the wall of pressurizer with ROSA-IV/LSTF experiment results indicated that the turbulent flow calculation results were more realistic. It was therefore concluded that the turbulent flow calculation was more reasonable and the possibility of migration of air to the hot leg was low in a pressurizer during reflux cooling.     

Fracture mechanics in creep situation interpretation of results on a test specimen

Fracture mechanics in creep situation is a difficult challenge for the 1990s. In France, CEA Saclay has conducted experimental tests on compact tension (CT) specimens at 650°C in order to investigate crack initiation under creep situations. The constitutive material is the 316SPH austenitic stainless steel used for most LMFR structures.Numerical simulations using SYSTUS code and simplified method analysis were performed on one of the tests (CT specimen at 650°C under constant load) to compare some parameters (notch opening, initiation time) with experimental values. The material constitutive law was represented by the complete elasto-viscoplastic CHABOCHE model for computation. Owing to geometrical characteristics such as thickness, the situation of the CT specimen was likely to be intermediate between plane stress and plane strain assumptions. From C^* parameter, incubation time obtained using the R5 rule was conservative in comparison with the test result.The continuum damage model developed at Ecole des Mines de Paris has also been used to assess creep damage in the notch tip area. The crack initiation time has been deduced from critical damage at characteristic distance (X_c = 0.05 mm). Considering critical damage specifically, for a CT specimen (D_c = 0.05), initiation time obtained was higher than the test result.The results of this study will contribute to the development of a methodology for nocivity analysis of cracks in creep situation.     

Thermal-hydraulic analysis on the effect of operator action in the loss of shutdown cooling system accident in a PWR

It is necessary to develop PSA methodology and integrated accident management technology during low power/shutdown operations. To develop this technology, thermal-hydraulic analysis is necessarily required to access the trend of plant process parameters and operator's grace time after initiation of the accident. In this study, the thermal-hydraulic behavior in the loss of shutdown cooling system accident during low power/shutdown operations at the Korean standard nuclear power plant was analyzed using the best-estimate thermal-hydraulic analysis code, MARS2.1. The effects of operator's action and initiation of accident mitigation system, such as safety injection and gravity feed on mitigation of the accident during shutdown operations are also analyzed.When steam generators are unavailable or vent paths with large cross-sectional area are open in the accident, the core damage occurs earlier than the cases of steam generators available or vent paths with small cross-sectional area. If an operator takes an action to mitigate the accident, the accident can be mitigated considerably. To mitigate the accident, high-pressure safety injection is more effective in POS4B and gravity feed is more effective in POS5. The results of this study can contribute to the plant safety improvement because those can provide the time for an operator to take an action to mitigate the accident by providing quantitative time of core damage. The results of this study can also provide information in developing operating procedure and accident management technology.     

Experimental determination of the boron concentration distribution in the primary circuit of a PWR after a postulated cold leg small break loss-of-coolant-accident with cold leg safety injection

It is known that under-borated coolant can accumulate in the loops and that it can be transported towards the reactor core during a loss-of-coolant-accident. Therefore, the mixing of weakly borated water inside the reactor pressure vessel was investigated using the ROCOM test facility. Wire-mesh sensors based on electrical conductivity measurement are used to measure in detail the spreading of a tracer solution in the facility. The mixing in the downcomer was observed with a measuring grid of 64 azimuthal and 32 vertical positions. The resulting distribution of the boron concentration at the core inlet was measured with a sensor integrated into the lower core support plate providing one measurement position at the entry into each fuel assembly.

The boundary conditions for this mixing experiment are taken from an experiment at the thermal hydraulic test facility PKL operated by AREVA Germany. The slugs, which have a lower density, accumulate in the upper part of the downcomer after entering the vessel. The ECC water injected into the reactor pressure vessel falls almost straight down through this weakly borated water layer and accelerates as it drops over the height of the downcomer. On the outer sides of the ECC streak, lower borated coolant admixes and flows together with the ECC water downwards. This has been found to be the only mechanism of transporting the lower borated water into the lower plenum. In the core inlet plane, a reduced boron concentration is detected only in the outer reaches of the core inlet. The minimum instantaneous boron concentration that was measured at a single fuel element inlet was found to be 66.3% of the initial 2500 ppm.     

Intermittent increase of electron current on a probe during a discharge in a magnetic field

An account is given of preliminary results in a study of the influence of a magnetic field on electron diffusion in a plasma. An intermittent increase was observed in the ratio of the electron current in the probe to the ion current at some critical value of the magnetic field intensity. According to the preliminary data, the critical magnetic field changes proportionally to the gas pressure. These facts evidently indicate the existence of two qualitatively different mechanisms for the transverse movement of the electrons, one of which is diffusion by collision.In conclusion, I wish to thank P. M. Morozov for his assistance and interest in this work, and also S. Sinotov, for taking part in the measurements.     

Numerical analysis and visualization experiment on behavior of borated water during MSLB with RCP running mode in an advanced reactor

The core bypass phenomenon of borated water injected through direct vessel injection (DVI) nozzles in APR1400 (Advanced Power Reactor 1400MWe) during main steam line break (MSLB) accidents with a reactor coolant pump (RCP) running mode has been simulated using a two-channel and one-dimensional system analysis model code (MARS), and a three-dimensional computational fluid dynamics (CFD) code (FLUENT). A visualization experiment has also been performed using a scaled-down model of the APR1400. The MARS analysis has predicted a serious core bypass phenomenon of borated water, while the CFD analysis has shown results opposite to the MARS results. The CFD analysis has shown that the flow pattern in the downcomer is fully three-dimensional and that vortex flow structures are formed near the cold legs so that the borated water might pass without difficulty into the high flow region of the cold legs and flow well into the lower downcomer. The visualization experiment has shown that the borated water flows well to the lower plenum, as in the CFD analysis. Both the CFD analysis and visualization experiment have proved that a serious core bypass phenomenon of borated water might not happen in the APR1400. These results are quite different from those predicted by MARS.     

Some insights into the role of axial gas flow in fuel rod behaviour during the LOCA based on Halden tests and calculations with the FALCON-PSI code

A model for axial gas flow in a fuel rod during the LOCA is integrated into the FRELAX model that deals with the thermal behaviour and fuel relocation in the fuel rods of the Halden LOCA test series. The first verification was carried out using the experimental data for the inner pressure during the gas outflow after cladding rupture in tests 3, 4 and 5. Furthermore, the modified FRELAX model is implicitly coupled to the FALCON fuel behaviour code.The analysis with the new methodology shows that the dynamics of axial gas-flow along the rod and through the cladding rupture can have a strong influence on the fuel rod behaviour. Specifically, a delayed axial gas redistribution during the heat-up phase of the LOCA can result in a drop of local pressure in the ballooned area, which is eventually able to affect the cladding burst. The results of the new model seem to be useful when analysing some of the Halden LOCA tests (showing considerable fuel relocation) and selected cases of LOCA in full-length fuel rods. While the short rods used in the Halden tests only show a very small effect of the delayed gas redistribution during the clad ballooning, such an effect is predicted to be significant in the full-scale rods - with a power peak located sufficiently away from the plenum - resulting in a considerable delay of the predicted moment of cladding rupture.