圆管通道底部再淹没过程实验研究

为验证和优化再淹没模型,通过实验研究了圆管通道内再淹没阶段流动换热特性,获得了不同工况下壁面温度的变化规律,实验工况范围为：入口冷却剂流速3~15 cm/s、入口过冷度15~75 ℃、初始壁面峰值温度340~600 ℃、实验压力0.2~0.4 MPa、加热功率1.3~2.3 kW/m。分析了初始壁温、冷却剂入口温度、入口流速及加热功率对骤冷时刻与骤冷温度的影响。结果表明,骤冷时刻与骤冷温度均随初始壁温、冷却剂入口温度以及加热功率的增加而增加,随入口冷却剂流速的增加而减小。     

Experimental observation of the droplet size change across a wet grid spacer in a 6 × 6 rod bundle

During the reflood phase of a postulated loss of coolant accident in a nuclear reactor, entrainment of liquid droplets can occur at a quench front of reflooding water. It is widely recognized that the behavior of the entrained droplets crucially affects the reflood heat transfer phenomena by decreasing the superheated steam temperature and interacting with a rod bundle and spacer grids. For this reason, various experimental and numerical studies have been performed to examine droplet behavior such as the droplet size, velocity and droplet fraction inside a rod array. In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle geometry, which results in the change of the interfacial heat transfer between droplets and superheated steam. A 6 × 6 rod bundle test facility in Korea Atomic Energy Research Institute was used for the experiment. Steam was supplied by an external boiler into the bottom of the test channel, and a droplet injection nozzle was equipped instead of simulating a quench front of reflooding water. The major measuring parameters of the experiment were the droplet size and velocity, which were measured by a high-speed camera and a digital image processing technique. A series of experiments were conducted with various flow conditions of a steam injection velocity, heater temperature, droplet size, and droplet flow rate. The experiments provided the data on the change of the Sauter mean diameter of droplets after collision with a wet grid spacer depending on flow conditions.     

A separate effect test on the cooling behavior in a pressurized water reactor core under a low reflooding rate condition by using ATLAS and RELAP5/MOD3.3 code assessment

A separate effect test was performed on the cooling behavior in a PWR core under a low reflooding rate condition by using the ATLAS (Advanced Thermal–Hydraulic Test Loop for Accident Simulation) which is a thermal–hydraulic integral effect test facility for the pressurized water reactors APR1400 and OPR1000. Although several integral tests for the reflood phase of a large break loss of coolant accident (LBLOCA) of APR1400 have been performed with the ATLAS, the previous integral effect tests for the reflood phase of a LBLOCA are not easily simulated by existing codes, such as the RELAP5/MOD3, due to a unique phenomena in ATLAS, that resulted from an injection of large amount of subcooled water onto the heated wall of which temperature was higher than the target value.     

Major findings from LBLOCA reflood tests using the ATLAS facility

A set of LBLOCA (large-break loss of coolant accident) reflood tests was performed in the first phase of the ATLAS (advanced thermal-hydraulic test loop for accident simulation) program. Their main objectives were to identify the major thermal-hydraulic characteristics during the reflood phase of a LBLOCA for APR1400 and to provide qualified data for APR1400 licensing. The ATLAS reflood test program could be divided into two phases (Phase-1 and Phase-2) according to the target period to be simulated. The Phase-1 tests were parametric effect tests for downcomer boiling in the late reflood phase of LBLOCA and the Phase-2 tests were integral effect tests for the entire reflood phase of LBLOCA. The experimental results from both Phase-1 and Phase-2 tests reproduced typical thermal-hydraulic trends expected to occur during the APR1400 LBLOCA scenario. A separate effect test was also performed under a low reflooding rate condition to provide data to validate the RELAP5 reflood models, and its experimental results showed a gradual reflooding in the core, a subsequent quenching of the core heater rods and the cooling of the reactor pressure vessel downcomer.     

百万千瓦级压水堆严重事故后再注水的有效性评价

根据现有的设计资料,使用一体化严重事故分析程序MELCOR1.8.6建立了核电厂一、二回路系统,非能动堆芯冷却系统和安全壳系统的模型,并模拟冷段2英寸(5.08cm)小破口叠加重力注入失效的严重事故发生后,将冷却剂注入堆芯的情形,分析其对严重事故进程的缓解能力。本文选取3个严重事故的不同阶段,将冷却剂分别以小流量(10kg/s)、中流量(50kg/s)和大流量(200kg/s)的速率注入堆芯,通过比较氢气产生量、堆芯放射性产生量及堆芯温度等数据来评估在严重事故不同阶段再注水的可行性。结果表明:在堆芯损伤初期,可认为10kg/s以上的流量足以冷却百万千瓦级事故安全。而当严重事故发展到堆芯开始坍塌阶段,200kg/s的注水流量可认为是基本可行的,而小于此流量的注水应慎重考虑。     

LOCA analysis of reactor cores with tight lattice using the modified RELAP5 code

The previous paper analyzed the reflooding phase of reactor cores with tight lattice. Models calculating the wall to fluid heat transfer in the precursory cooling region and in the vicinity of the quench front were developed and validated in the previous paper (Wu et al., 2012). In this paper, these newly developed models were used to modify RELAP5/MOD3.2 in order to make the code be suitable for tight lattice. Besides, minor modifications to the wall friction model and bubbly-slug interfacial drag model were done. Then the newly developed code RELAP5/MOD3.2/TIGHT was used to analyze the LOCA transients of conceptually designed reactor cores with three types of tight lattice. The results showed that the peak cladding temperatures in the reflooding phase are much higher than that in the blow-down phase. Through comparison between the calculation results of LOCA transients of the three types of tight lattice, it was found that with smaller pitch to diameter ratio, the peak cladding temperature was much higher. LPIS injection flow rate should be increased in order to keep the rod cladding temperature be within the LOCA criteria. Steam generation will prevent the coolant from flowing downstream of the channel in reactor cores with a very small flow area. From the reactor safety aspect and the economic aspect, we do not recommend that reactor cores be designed with p/d ratio less than 1.10.     

Onset Criterion for Liquid Entrainment in Reflooding Phase of LOCA

Reflooding tests were conducted in a rod bundle geometry at the maximum pres- sure of 12 MPa to investigate thermal-hydraulic behavior during a small break loss-of-coolant accident (SBLOCA) in a nuclear reactor. The test conditions ranged 0.6 ～ 12 MPa for pressure, up to 920 K for rod surface temperature, up to 20 cm/s for bundle inlet flow velocity and up to 2 kW/m for linear power input. The principal objective of this paper is to investigate the onset condition for liquid entrainment by steam flow in the relatively high pressure reflooding phase. Experimental results showed a tendency that the liquid entrainment decreased with increase in pressure when the other parameters such as an inlet flow rate and rod temperature were fixed. A new correlation for the onset criterion for liquid entrainment was derived from the experimental results and an analysis of a force balance for a liquid droplet. Effects of pressure on liquid entrainment in the reflooding phase were made clear from the experimental and analytical results.     

竖直管道内间歇式两相流动沸腾特性分析

自然循环或重力注水过程的热功率、冷却剂流量等操作条件较小,易出现各种流动不稳定现象,影响核反应堆事故的发展进程,间歇式流动沸腾现象就属于其中的一种。以去离子水为工质,采用2×2加热棒束,对内径为32 mm竖直通道内的间歇式流动沸腾现象进行了实验研究,分析了不同热流密度下间歇式流动沸腾不稳定现象的变化规律,讨论了热流密度对间歇式沸腾周期的影响。结果表明,在一定的热流密度条件下,当加热通道内流体达到饱和并过热时,会发生周期性地剧烈喷涌及冷液回流现象,期间伴随泡状流、弹状流、搅混流及环状流等多种流动形态;间歇喷涌周期取决于沸腾停滞时间,随热流密度的不断增大,沸腾停滞时间缩短,间歇喷涌周期也缩短。当热流密度增大到一定程度时,间歇式流动沸腾现象消失,从而转变为另一种两相流动不稳定现象。     

矩形窄缝通道底部再淹没实验研究

为了解矩形窄缝通道在失水事故（LOCA）下底部再淹没过程中的热工水力特性,在不同实验条件下开展再淹没实验研究。矩形窄缝通道由2块因科镍合金焊接而成,本研究根据温度变化曲线分析底部再淹没过程,计算并对比不同实验工况下的骤冷前沿的推进速度（骤冷速度）,以及研究实验参数对再淹没过程的影响。实验结果表明,底部再淹没骤冷速度随着系统压力增大、进口流速增大、初始壁面温度降低以及冷却水过冷度的增大而增大。对比分析底部与联合再淹没工况,结果表明流量相同的情况下,底部再淹没的骤冷速度大于联合再淹没。本文研究为板状燃料元件反应堆事故预防以及事故缓解等研究奠定了基础。      

水平加速条件下矩形通道内冷却剂温度与流场分析

运用流体计算软件模拟计算和分析了处于船用反应堆某处的矩形冷却剂通道在随船体水平加速运动时冷却剂的温度和流场，考虑了由于船体加速运动而引起的流量孔板流量分配的变化对矩形通道入口速度的影响。计算结果表明，在船体的水平加速度较大时，通道出口冷却剂的温度与流场的变化很大，但持续时间较短。     

Natural circulation flow behavior at reduced inventory in a VVER geometry

This paper deals with the natural circulation flow characteristics of the VVER-440 geometry at reduced coolant inventory. Special emphasis is on the flow rate of the primary circuits during the two-phase flow regime. For studying two-phase natural circulation flow phenomena in a VVER geometry a series of cold leg small break loss-of-coolant accident (SBLOCA) tests was carried out in the PArallel Channel TEst Loop (PACTEL), a 1/305 volumetrically scaled model of a VVER-440 reactor. The tests were conducted with break areas ranging from 0.1 to 1.5 % of the scaled cold leg cross-sectional area of the reference reactor. A partial failure of the high-pressure injection system (HPIS) was assumed. The tests reveal a trend towards an increasing primary circuit mass flow rate with decreasing inventory. This contradicts the findings of earlier tests in multi-loop VVER geometry. With single-loop facilities, increased mass flow rates at reduced inventories have been reported before. The increase of the two-phase flow rate turns out to be a consequence of the combined effect of break size, pressure range and secondary side feed and bleed procedure. The physical phenomena of flow stagnation in the primary circuits, system pressurization, asymmetric loop flows, and loop seal clearing and refilling take place during the natural circulation cooling process from single-phase into two-phase and boiler–condenser modes. In addition, flow reversal in the undermost tubes of the horizontal steam generators (SG) is observed. These phenomena are discussed briefly while a general insight into the course of the tests is presented.     

Cladding tube deformation and core emergency cooling in a loss of coolant accident of a pressurized water reactor

The paper summarizes the dominant effects which finally ensure the core coolability of a pressurized water reactor in a loss-of-coolant accident (LOCA).The main results are summarized as follows:

• — The cooling effect of the two-phase mixture which is intensified during reflooding increases temperature differences on the cladding tube circumference and thus limits the mean circumferential burst strains to values of about 50%.

• — An unidirected flow through the fuel rod bundle during the refill and reflooding phases causes maximum cooling channel blockage of about 70%.

• — The coolability of deformed fuel elements can be maintained up to flow blockages of about 90%.

All effects investigated indicate that in a LOCA no impairment of core coolability and public safety has to be expected.     

Transient tests on blower trip and rod removal at the HTR-10

Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes.Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results.Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 °C during two tests.The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW.The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of $ 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.     

Assessment of the models in RELAP/SCDAPSIM with QUENCH-06 analysis

A fundamental principle of accident management in a nuclear power plant is the injection of water to cool the core. In this framework, a series of QUENCH tests have been conducted at Karlsruhe Institute of Technology (formerly Forschungszentrum Karlsruhe). The test results constitute a significant experimental database not only for further understanding of reflooding behavior, but also for code validation and improvement. The RELAP/SCDAPSIM code is a system code that is used to model reactor behavior and is widely used around the world. To date, assessment and validation have been performed with numerous experiments, including QUENCH tests. In the previous studies, the results of QUENCH simulations were referred to be sensitive to two main parameters: the electrical resistance and the thermal conductivity of the shroud insulator, which are subject to relatively large uncertainty. It is important to investigate these two parameters in detail, because this would enable identification of those SCDAP models that require further improvement. In this study, the uncertainty of the electrical resistance was reduced by modification of the code and subsequent validation with experimental data. In addition, modification of the thermal properties of the shroud insulator is suggested with consideration of the argon atmosphere in the facility. Finally, upcoming problems and questions are discussed. A rather good agreement was obtained than those of previous studies. As a result, more accurate modeling of the electrical resistance and the thermal properties of the shroud insulator was conducted and the importance of these parameters was evaluated.     

环形通道内再淹没过程骤冷前沿推进速度实验研究

骤冷前沿推进速度是衡量失水事故中再淹没过程堆芯冷却效率的重要参数之一。本文通过实验研究了竖直环形通道内骤冷前沿的推进特性,获得初始壁温、入口温度、入口质量流速及加热功率对骤冷前沿推进速度的影响。实验结果表明,骤冷前沿推进速度随初始壁温、入口温度和加热功率的增加而减小,随入口质量流速的增加而增加。      

A study of the unsteady flow and heat transfer in the reflooding of water reactor cores

In the event of a loss-of-coolant accident in a water-cooled reactor, the primary consideration is terminating the clad temperature excursion caused by release of the stored and decay heat in the fuel. This requires that emergency coolant injection systems reflood the reactor core.For certain break positions, the pressure loss incurred by venting steam partially offsets the hydrostatic head available to drive flow through the core. Flow oscillations can also be set up due to the fluid inertia and vapour compressibility.The present paper reports the results of an extensive series of experiments performed on unstable reflooding, covering wall temperatures up to 1000°C and reflooding rates typical of reactor values. Measurements are reported of quenching rates, oscillation frequencies and pre-quench heat transfer.It is shown, except for a short initial period of violent oscillations, that the rewetting rate and pre-quench heat transfer, for a given mass flow rate, are relatively unaffected by the presence of oscillations. The average pre-quench heat transfer coefficient is shown to vary as (water mass flow rate)ⁿ where n = 0.5–0.7, consistent with available world data.Theory and experiments also show that there is a critical value of outlet loss coefficient, for a given power level, where no further advance of the quench front can occur, the back pressure completely offsetting the available driving head for core reflooding. This value is much greater than the outlet loss coefficient for typical reactor designs, thus ensuring core reflooding. The critical loss coefficient is suggested as the relevant parameter for scaling purposes.A new theoretical model for the oscillations is derived which is shown to predict the oscillation frequencies of all available data. It is also shown that the frequency and damping are only weakly dependent on: upper plenum flow area, size of vapour space, effective inertia of water oscillating and pressure, and are independent of the outlet loss coefficient.     

SCC testing of pipe materials in BWR environment

A test loop has been installed in Ringhals 1 BWR, including facilities for Constant Elongation Rate Testing (CERT) and Electrochemical Potential (ECP) measurements in primary reactor water at reactor operation temperature. The loop is designed as to minimize transport time for reactor water from the reactor pressure vessel to the specimens being tested. Thus the testing environment is representative of the primary piping systems of BWRs, also with regard to short-lived constituents like hydrogen peroxide.The test program, which is in progress, has covered seven tests during start-up conditions or during power operation with presently current reactor water chemistry. In this presentation only CERT testing results on heavily sensitized austenitic chromium—nickel stainless steel are presented, although many other materials have been tested.Results show sensitized austenitic stainless steel is more prone to intergranular stress corrosion cracking (IGSCC) in actual than in simulated BWR environment and that start-up environment is chemically more aggressive than power operation environment. Reproducibility of the CERT technique as used is excellent.     

Thermal and hydraulic tests of test sections in the helium engineering demonstration loop

The helium engineering demonstration loop (HENDEL) has been constructed and operated to test the large-scale components of the high temperature engineering test reactor (HTTR) under simulated reactor operating conditions. The fuel stack test section (T₁) of HENDEL simulates the fuel stack of the HTTR core and is used to investigate thermal and hydraulic performance. Hot tests with 1000°C helium gas have been conducted using simulated fuel rods having uniform, exponential and cosine axial heat flux distributions. The test results agreed with previously proposed correlations, although the simulated fuel rods had various heat flux distributions and high heat flux rates.

The in-core structure test section (T₂) also was installed in the HENDEL to verify the performance of the core bottom structure of the HTTR. The tests show that good performance was obtained. Examination of the thermal mixing characteristics indicated that mixing started at the location where the hot helium gas flowed into the hot plenum and that complete mixing was achieved during the downward flow in the outlet hot gas duct. The seal performance testing indicated no change of the leakage flow rate after 4000 hours of operation. The temperature of the metal portion of the structure was below 500°C and uniform around circumferential cross-sections due to the good performance of the thermal insulation blocks.     

基于ECC安注热混合试验的LOCUST 1.2分析与验证

在失水事故（LOCA）工况下安注系统投入使用时,蒸汽与安注冷却剂会发生流体热力学混合,热混合过程中冷腿段的冷却是直接影响堆芯再淹没与否的重要因素。中国广核集团有限公司自主研发了一款两相流热工水力系统分析软件LOCUST,可用于压水堆核电厂事故工况的分析计算。基于西安交通大学堆芯应急冷却系统（ECCS-XJTU）试验台架进行的堆芯应急冷却（ECC）安注热混合试验,本文使用LOCUST软件对ECC热混合试验进行了几何建模及计算分析。ECC热混合试验工况主要为不同流量下主管纯蒸汽与安注管过冷水的混合,蒸汽流量为25~125 kg/h,过冷水流量为100~500 kg/h。模拟计算结果和试验结果的对比分析表明：试验段出口质量流量计算值的最大相对误差在13.8%以内,混合后温度计算值的最大相对误差在8%以内,LOCUST在计算高温蒸汽和过冷水混合时的计算结果相对保守,总体上验证了LOCUST在LOCA下两相热混合安注计算的可靠性和准确性。     

Demonstration of the heat removing capabilities of the EPR core catcher

The EPR implements an additional, fourth level of defense-in-depth that aims at limiting and restricting the consequences of a postulated severe accident with core melting to the immediate vicinity of the plant. As this requires an intact confinement, it is necessary, among others, to avoid an attack of the molten core on the basemat. For that purpose, the EPR includes a large ex-vessel core catcher. It increases the surface-to-volume ratio of the melt after its release from the reactor pressure vessel (RPV) and allows the effective quenching and stabilization of the melt before it can attack the structural concrete.The bottom and sides of the core catcher are cooled by a system of horizontal water-filled channels. The water is provided either passively, by overflow from an internal reservoir, or actively by the containment heat removal system (CHRS).To quantify the heat removing capability of the horizontal part of the proposed cooling structure, a set of experiments in a full-scale, horizontal, 5 m long cooling channel have been performed. To simulate decay heat, the channel was electrically heated from the top. The experiment was integrated in the BENSON test rig, a highly flexible, separate-effect test facility operated by Framatome ANP. In accordance with the potential later modes of operation, both co-current and counter-current flow of the water/steam mixture have been investigated.The tests demonstrated the good-natured behavior of the system, even for induced heat fluxes that significantly exceed realistically expectable maximum values. Although, at high heat fluxes, a local dry-out occurred at the top of the channel, structural temperatures remained in a safe range. This excellent performance is attributed to the fact that heat can enter the water through both the horizontal and vertical surfaces of the cooling channel. As a result a high, effective critical heat flux (CHF) level is achieved. The performed tests yield a valuable contribution to the validation of the function of the EPR core catcher concept.