600MW核电站控制棒驱动机构原理样机风冷试验

描述了秦山核电二期区动机构原理样样风冷及发热试验。通过模拟反应堆实际运行条件,确定了驱动机冷却风道的压力损失,所需的最小风冷速度以及驱机构发热量与风冷速度的关系,明确了驱动控制的运行要求。     

Calculation of Iodine Removal by Spray in LWRs Containment Vessels

A computer code MIRA-PB for predicting the iodine removal by containment spray in LOCA was prepared on the basis of MIRA-P/MIRA-B code developed in Battelle Columbus Laboratories. MIRA-PB considers behavior of inorganic iodine, organic iodide, and iodic aerosol and simultaneous removal by natural deposition, liquid-film absorption, spray washout, filtration and leakage to the environment. The iodine removal by the containment spray systems in LOCA of PWR and BWR is calculated with the MIRA-PB.     

Development of Evaluation Code for Long-Term Cooling of Containment with Passive Safety Feature

Simplified BWRs are characterized as an adoption of a passive ECCS and a passive containment cooling system (PCCS). While a passive ECCS has a short term core cooling function, a PCCS has a long-term decay heat removal function. As a PCCS, several concepts, differing in cooling location and method employed, have been considered. From the containment thermal- hydraulic response analysis viewpoint, simplified BWRs are essentially different from the current BWRs. For evaluating and comparing the performance of several PCCSs over full break spectra, the new containment safety evaluation code TOSPAC was developed as a preliminary design tool for PCCS. This paper summarizes the thermal-hydraulic modelings of the TOSPAC code and the validity evaluation of the TOSPAC code, compared with TRAC-BF1 calculation.

From the validity evaluation concerning a main steam line break (MSLB) accident analysis for an isolation condenser (I/C) as a PCCS, it was found that the TOSPAC calculation result shows reasonable agreement with that for TRAC, even though the TOSPAC consists of simpler modelings.     

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.     

软X射线诊断冷却测试平台设计

软X射线物理诊断普遍应用于托卡马克聚变装置。在实验时,上限温度仅为75℃的软X射线诊断探测器将处于250℃的环境温度中,因此冷却系统不可或缺。本文分析了软X射线诊断探测器在运行时将受到的热负荷功率,并依据计算结果及实验时的冷却条件,创造性的设计了采用水、氦气混合制冷方式的冷却测试平台。所进行的测试结果表明,平台运行稳定,不仅能够很好的满足软X射线诊断的冷却需求,同时也为其它聚变诊断冷却系统的设计提供借鉴。     

国内压水堆核电厂安全壳喷淋试验的比较分析

以秦山第二核电厂1、2号机组和田湾核电厂1、2号机组为例,对国内已运行的压水堆核电厂在调试阶段进行安全壳喷淋试验时所采用的两种方式进行了比较分析,总结出两种喷淋试验方法各自的优缺点,可为国内新建核电机组的安全壳喷淋系统的调试试验提供借鉴。     

Assessment of SAFER03 Code Using ROSA-III Break Area Spectrum Tests on Boiling Water Reactor Loss-of-Coolant Accidents

The SAFER03 computer code has a newly developed evaluation model for the analysis of various boiling water reactor (BWR) loss-of-coolant accidents (LOCAs). Analyses of the ROSA-III break area spectrum tests in a recirculation line were performed using the SAFER03 to assess the predictive capability of the code for a BWR LOCA. The ROSA-III test facility at the Japan Atomic Energy Research Institute (JAERI) was constructed to simulate a LOCA in a BWR/6-251 plant with 848 fuel bundles and 24 jet pumps. This paper summarizes the assessment results of SAFER03 which predicted the system responses and key phenomena well and the conservative peak cladding temperature (PCT) for recirculation line break tests with different break areas.     

非能动堆芯冷却系统LOCA下冷却能力分析

本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统（PXS）中非能动余热排出系统（PRHRS）、堆芯补水箱（CMT）、安注箱（ACC）、自动卸压系统（ADS）和安全壳内置换料水箱（IRWST）等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1 477 K,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。     

Initial Thermal-Hydraulic Behaviors under Simultaneous ECC Water Injection into Cold Leg and Upper Plenum in a PWR-LOCA

Core thermo-hydrodynamic characteristics under the combined injection mode before and just after the beginning of bottom reflood of a PWR-LOCA were experimentally studied by performing three tests in Slab Core Test Facility simulating a full radius slab section of a PWR. Emergency core cooling water was simultaneously injected into the upper plenum and the intact cold leg. The subcooling and the radial distribution of the upper plenum injection water were the test parameters.

The core was cooled by falling water before the beginning of bottom reflood. However, the core was finally quenched by bottom reflood. Before the beginning of bottom reflood, the transients of water level in the lower plenum were different among three cases, that is, the water level was rapidly or gradually increased in the first and second cases, respectively, or remained below the bottom of core barrel in the third case. The bottom reflood was much delayed in the last two cases. Even under the conditions with large upper plenum injection rate of subcooled water and with steam escape through the lower plenum, continuous fall back was not observed but the subcooled water was intermittently supported by the upward steam flow generated in the core.     

Restrictive Effect of Ascending Steam on Falling Water during Top Spray Emergency Core Cooling

Water spraying experiments were conducted to find out a flow rate of falling water overcoming ascending steam during top spray emergency cooling with an 8×8 type simulated fuel rod bundle of real size. The bundle consisted of 64 rods, each with a diameter of 12.5 mm, arranged in the form of square lattice with a pitch of 16.3 mm. In the experiments the simulated fuel rods were not heated. Instead, steam was injected into the lower plenum vessel simulating bundle-generated steam. As the results, (1) a criterion was proposed to determine the region where the restrictive effect of ascending steam on falling water appears, considering the decrease of a flow rate of ascending steam due to condensation by a spray of subcooled water, (2) the restrictive effect was independent of water head on the upper tie plate and water injection methods, and (3) an analytical model based on the pressure balance at the upper tie plate was proposed to calculate a flow rate of falling water overcoming ascending steam.     

Analysis of Precursory Cooling in Quenching Phenomena

It has been noted that precursory cooling plays an important role in quenching phenomena. In this work, a new model is presented by assuming that the heat transfer in precursory cooling is mainly due to film boiling, which persists in a finite length. Then the quench velocity and temperature profile are obtained based on the three-region model for one- dimensional axial heat conduction in a heated tube. We applied our model to several existing experimental results and obtained a correlation to predict the effective length of precursory cooling region. It turns out that the correlation takes an identical form for both falling-film rewetting and bottom flooding.     

New Non-Equilibrium Thermal-Hydraulic Model,Theory

Corresponding with each non-equilibrium state given by the pressure p, the specific enthalpy of the two-phase mixture h and the void faction α, an equilibrium state defined by p and h is assumed to exist. The relaxation equation for α which relates the non-equilibrium state to the corresponding equilibrium state is derived from the mass and energy equations for the gas phase. The conservation equations for the mixture are spatially differenced in such a way that direct node-node coupling can not take place. Then, the space differencing scheme brings about the three step reduction of the conservation equations, for each of which the physical interpretation is possible. The hydraulic boundary conditions as well as the pump and valve behaviors are incorporated in the model in a natural manner. The model has been implemented in the THYDE-P code to analyze the various dynamic behaviors of the coolant systems.     

Effect of Gas Atoms on Recovery of Work Hardening in Al-Li Alloys

In the first report of this study, dealing with CCFL and CCFL breakdown phenomena associated with the injection of emergency core cooling spray water into upper plenum during refill-reflood phase of a BWR LOCA, the following tests results were obtained.

The injected water maintained two-phase pool across the top of entire core after CCFL breakdown. The pool level oscillated near spray elevation. The objective of this paper is to clarify the mechanism of these phenomena, evaluating steam and spray flow effects on CCFL breakdown.

It is found that when spray flow rate was slightly larger than the CCFL drainage deter- mined by core steam flow, pool maintained at some constant level near spray elevation, after CCFL breakdown. On the other hand, when spray flow was appreciably larger than CCFL drainage, pool level slowly oscillated. The oscillation was caused by significant changes in steam condensation rate, and the corresponding subcooling penetration into the fuel bundles, when the pool level passed the spray elevation. The TRAC-BD1 analysis of test results suggested the small sector wall effect of test apparatus on CCFL breakdown phenomena.     

Analytical Study of Thermo-Hydrodynamic Behavior of Reflood-Phase during LOCA

The objective of this study is the establishment of the thermo-hydrodynamic model of the reactor core during reflood phase of LOCA.

Based on the quench model proposed by the author, and assuming a reflood model including a flow model and a set of the thermo-hydrodynamic correlations, a reflood analysis code named “REFLA-1D” was developed.

Considerably close agreement between PWR-FLECHT tests and the results calculated by REFLA-1D code for the critical Weber number We_c= 1 was obtained for fuel clad temperature histories and the quench time and the quench temperature except for the quenching from the top of the fuel rod. It was found that the errors of calculated quench time and temperature are within ±20% under the following conditions: (1) pressure 4.5–1.5 kg/cm²·a or core inlet velocity 15–4.8 cm/s, (2) inlet subcooling more than 30°C. In the transition flow region, the calculated tendency of the temperature histories is different from the measured. This reflood model appears to be reasonable but some modifications on the low flooding quench model and the transition flow are necessary.     

PRHRS隔离阀前后破口事故对非能动堆芯冷却系统的影响分析

对于AP型核电站小破口失水事故(SBLOCA)试验进程,国内外有较为一致的认识,但对于相同尺寸破口在不同破口位置对试验进程、非能动堆芯冷却系统的影响仍需进一步研究。本文利用大型非能动堆芯冷却整体试验台架ACME开展了非能动余热排出系统（PRHRS）隔离阀前后破口事故试验工况研究,并以堆芯补水箱（CMT）侧冷管底部破口事故工况作为对比工况。试验结果表明：ACME开展的PRHRS隔离阀前后破口事故模拟工况事故进程符合典型SBLOCA进程,堆芯始终处在良好的冷却状态,非能动堆芯冷却系统的安全性得到有效验证;相同破口尺寸工况下,不同破口位置对事故进程有一定的影响,其中破口位置对CMT液位、安注流量的影响较为关键。对比工况中,PRHRS设备换热量也有较大不同,冷管破口和隔离阀后破口工况较隔离阀前破口工况换热量更大,但PRHRS换热管内部流动换热机理需进一步研究。     

Overview of SAFER03 Assessment Studies

The safety research for BWRs has been positively done by the JAERI, Japanese BWR utilities and BWR vendors in this decade and has shown the important phenomena under BWR LOCA conditions. Based on these significant results, the SAFER03 computer code was jointly developed by Toshiba, Hitachi and General Electric. SAFER03 has been qualified against the BWR simulation test data obtained from TBL, ROSA-III and FIST-ABWR test facilities. The objectives of this study are to assess the predictive capability of SAFER03 code to simulate the significant LOCA phenomena and to catch key parameters during BWR LOCA. This paper summarized the results of these SAFER03 assessments and showed that SAFER03 could predict the realistic behavior of BWR LOCA with slight conservative peak cladding temperatures.     

Analysis of Effect of SBLOCA before and behind Isolation Valve of PRHRS on Passive Reactor Core Cooling System

For the test process of small break loss of coolant accident (SBLOCA) of AP type nuclear power plant, there is a more consistent understanding at home and abroad. However, the influence of the same size of the break on the test process and passive core cooling system in different locations still needs further study. In this paper, a large passive core cooling integrated test facility ACME was used to study the break accident test conditions of passive residual heat removal system (PRHRS) before and behind the isolation valve, and the bottom break test of the cold pipe of core makeup tank (CMT) was used as the contrast condition. The test results show that the accident process of PRHRS before and behind the isolation valve is in accordance with the process of SBLOCA, the core is always in a good cooling statement and the safety of passive core cooling system is effectively verified. There is a certain impact on the accident process for the same break size and different break locations, and the location of the break has a key impact on the CMT level and safety injection flow. In contrast, the heat transfer of PRHRS equipment is also quite different. The heat transfer of cold pipe break and break behind the isolation valve is greater than break before the isolation valve, however, the flow and heat transfer mechanism of PRHRS heat exchange tube needs further study.     

Boiling Water Reactor Loss of Coolant Tests Single Failure Tests with ROSA-III

The single failure tests with the ROSA-III were simulated BWR LOCA experiments by the scaled BWR test facility resulting from a 200% double-ended break at the recirculation pump suction line to evaluate the core cooling capability of a BWR ECCS under the single failure condition.

The experimental results showed that the loss of LPCS and one LPCI resulted in the highest PCT of 870 K of the single failure series tests, yet a core cooling capability by the ECCS was maintained. The REALP4/Mod 6 code was used to evaluate the predictive capability of the LOCA analysis code. The calculated results showed that the RELAP4/Mod 6 code was able to predict occurrences and sequence of major events anticipated to occur during a BWR LOCA correctly. However it was found that the code still needs to be improved in a CCFL model to better describe thermohydraulic behavior in the core.

The analyses presented in this paper are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict the system response of a BWR during a LOCA.     

Experimental Study on Heat Removal Characteristics for Water Wall Type Passive Containment Cooling System

To evaluate the heat removal capability of a water wall type cooling system, which is one passive containment cooling system (PCCS), the thermal hydraulic behavior in the suppression pool (S/P) and the outer pool (O/P, flat plate water wall) have been investigated experimentally. The following results were obtained. (1) A thermal stratification boundary, which separates the pools into the upper high temperature and lower low temperature regions, was formed just below the vent tube outlet. (2) Convection heat transfer characteristics in the S/P and O/P along the primary containment vessel (PCV) wall had no significant differences and were those of natural convection. Correlation of the natural convection heat transfer up to the Ra number of 2×10¹⁴ was obtained. (3) Vertical variations of local condensation heat transfer coefficients under a noncondensable gas presence were within ±10% of the average value for the 4.7 m heat transfer length. An experimental correlation for the average condensation heat transfer coefficients was obtained as a function of steam and noncondensable gas mass ratio. (4) An analytical model to evaluate the system performance of the water wall type PCCS was verified. (5) A baffle plate concept to mitigate thermal stratification at the vent outlet and to enlarge the high temperature region in the S/P was considered as a means to improve heat release capability. Thermal hydraulics with a baffle plate were examined, and effectiveness of the baffle plate to improve the heat release capability was confirmed.     

LOFT Experiment LP-02-6 Analysis by RELAP5/MOD2 Code with Improved Minimum Film Boiling Temperature

Groeneveld-Stewart's minimum film boiling temperature correlation was incorporated into the RELAP5/MOD2 code in order to explicitly define the minimum film boiling temperature. The transition boiling curve in the code was also modified. The Loss-of-Fluid Test (LOFT) experiment, Experiment LP-02-06 which was a cold-leg double-ended break LOCA experiment with minimum emergency core coolant injection, was analyzed with the modified RELAP5/MOD2 code. The modified RELAP5/MOD2 code well calculated system transients including the rod surface temperature transient. The temporary rewetting of rods in the early phase of blowdown, which had not been predicted by the original RELAP5/MOD2 and other codes, was predicted by the modified RELAP5/MOD2 code.