采用典型模糊控制器实现压水堆稳压器的综合控制

对压水堆稳压器的压力和水位控制．提出了一种模糊综合控制方案。采用3个典型模糊控制器分别对电加热器、喷淋卸压阀和上充阀进行控制;在稳压器压力典型模糊控制器中采用了积分分离方法。本文对汽轮机负荷阶跃变化、线性变化、甩负荷3种工况进行了控制系统的仿真实验。结果表明,稳压器的压力以及水位的瞬态和稳态控制性能都得到了较大改善,明显优于GA-FC和PID控制方案。     

Benchmark calculations of pressurizer model for Maanshan nuclear power plant using TRACE code

The pressurizer plays an important role in controlling the pressure of the primary coolant system in pressurized water reactor (PWR) nuclear power plants. An accurate modeling of the pressurizer is needed to determine the pressure response of the primary coolant system, and thus to successfully simulate overall PWR nuclear power plant behavior during transients. The purpose of this study is to develop a pressurizer model, and to assess its pressure transients using the TRACE code version 5.0. The benchmark of the pressurizer model was performed by comparing the simulation results with those from the tests at the Maanshan nuclear power plant. Four start-up tests of the Maanshan nuclear power plant are collected and simulated: (1) turbine trip test from 100% power (Test PAT-50); (2) large-load reduction at 100% power (Test PAT-49); (3) net-load trip at 100% power (Test PAT-51); and (4) net-load trip at 50% power (Test PAT-21). The simulation results show that the predictions of the pressure response are in reasonable agreement with the power plant's start-up tests, and thus the pressurizer model built in this study is successfully verified and validated.     

MAAP5 simulation of the PWR severe accident induced by pressurizer safety valve stuck-open accident

The Modular Accident Analysis Program version 5 (MAAP5) is a computer code that can simulate the response of light water reactor power plants during severe accident sequences. The present work aims to simulate the severe accident of a typical Chinese pressure water reactor (PWR) with MAAP5. The pressurizer safety valve stuck-open accident is essentially a small break loss-of-coolant accident (SBLOCA), which becomes one of the major concerns on core melt initiating events of the PWR. Six cases with different assumptions in the pressurizer (PZR) safety valves (SVs) stuck-open accident stuck open accident were analyzed for comparison. The results of first three cases show that the severe accident sequence is correlated with the number of the stuck open valve. The primary system depressurized faster in a more SVs stuck open case, and the consequences in which is hence slighter. The remaining 3 cases along with the case 2 were then analyzed to study the effect of operator intervention to the accident. The results show that the auxiliary feed water (AFW) is effective to delay the core degradation and hence delayed the finally system recovery. The high pressure injection (HPI) operation and manually opening the steam generator (SG) SVs are effective to mitigate this kind of severe accident. The results are meaningful and significant for comprehending the detailed process of PWR severe accident, which is the basic standard for establishing the severe accident management guidelines.     

Evaluation of intentional depressurization strategy in Chinese 600 MWe PWR NPP

Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization.     

The thermal-hydraulics of a simulated PWR facility during steam generator tube rupture transients

The thermal-hydraulics of the semi-scale test facility during steam generator tube rupture transients were investigated in this paper. The test facility simulates the main features of a Westinghouse four-loop pressurized water reactor (PWR) plant.The constructed analytical model simulated both the intact and broken loops, and included the vessel (lower plenum, core, upper plenum, upper dome), the hot legs, pressurizer and the primary and secondary sides of the U-tube steam generators. The two-phase Modular Modeling System code, which was developed by the Electric Power Research Institute, and the EASY5 simulation language were used in carrying out the calculations. A control model was developed to simulate the major facility control systems and to perform the necessary control functions.Calculations were carried out during the first three hundred seconds of the event, where the automatically functioning plant protection system components were assumed to operate. The impact of reactor scram, pressurizer heater activation, main steam isolation valve closure, emergency core cooling system activation, pump trip, main feedwater termination, auxiliary feedwater injection, and atmospheric dump/safety relief valves opening/closing on the system response was calculated.The time histories of the thermal-hydraulic conditions, such as pressure and temperature, are presented for one, five and ten-tube ruptures. Comparisons with experimental data and RELAP-5 (MOD 1.5) calculations are also given.     

稳压器模糊控制系统初步研究

通过合理地确定模糊集、隶属度函数和模糊控制规则等。初步研究和设计了核电站稳压器的模糊控制器。仿真结果表明,与传统的比例－积分－微分控制器相比,模糊控制器在提高稳压器压力、水位控制系统的动态和稳态性能方面具有明显的优点。     

Computational analysis for prediction of pressure of PWR presurizer under transient conditions

A computer model has been developed for prediction of the pressure in the pressurizer under transient conditions.In the model three separate thermodynamic regions which are not required to be in thermal equilibrium have been considered.The mathematical model derived from the general conservation equations includes all of the important thermal-hydraulics phenomena occurring in the pressurizer,i.e.,stratification of the hot water and incoming cold water,bulk flashing and condensation ,wall condensation,and interfacial heat and mass transfer,etc.The bubble rising and rain-out models are developed to describe bulk flashing and condensation.respectively.To obtain the wall condensation rate,a one-dimensional heat conduction equation is solved by the pivoting method.The presented model will predict the pressure-time behavior of a PWR pressurzer during a variety of transients.The results obtained from the propesed mathematical model are in good agreement with available data on the CHASHMA nuclear power plant‘s pressurizer performance.     

Verification of a TRACE EPR™ model on the basis of a scaling calculation of an SBLOCA ROSA test

In cooperation with the Finnish Radiation and Nuclear Safety Authority (STUK), a project has been launched at the Paul Scherrer Institute (PSI) aimed at performing safety evaluations of the Olkiluoto-3 nuclear power plant (NPP), the first EPR™, a generation III pressurizer water reactor (PWR); with particular emphasis on small-and large-break loss-of-coolant-accidents (SB/LB-LOCAs) and main steam-line breaks.As a first step of this work, the best estimate system code TRACE has been used to develop a model of Olkiluoto-3. In order to test the nodalization, a scaling calculation from the rig of safety assessment (ROSA) test facility has been performed. The ROSA large scale test facility (LSTF) was built to simulate Westinghouse design pressurized water reactors (PWR) with a four-loop configuration. Even though there are differences between the EPR™ and the Westinghouse designs, the number of similarities is large enough to carry out scaling calculations on SBLOCA and LOCA cases from the ROSA facility; as a matter of fact, the main differences are located in the secondary side. Test 6-1 of the ROSA 1 programme, an SBLOCA with the break situated in the upper head of the reactor pressure vessel (RPV), was of special interest since a very good agreement with the experiment was obtained with a TRACE input deck. In order to perform such scaling calculation, the set-points of the secondary relief and safety valves in the EPR™ nodalization had to be changed to those used in the ROSA facility, the break size and the core power had to be scaled by a factor of 60 (according to the core power and core volume) and the pumps coast down had to be adapted to the ones of the test. The calculation showed very similar results as the experiment and the ROSA-TRACE calculation. The only significant difference observed was a faster primary depressurization after the break flow turned to single-vapor flow. This difference could be explained on the basis of geometrical differences between the EPR™ and ROSA/Westinghouse RPV's designs.     

核电厂严重事故下卸压对氢气产生的影响分析

研究了1000MWe压水堆核电厂在典型的高压严重事故序列下卸压对氢气产生的影响。分析结果表明,开启1列、2列和3列卸压阀进行一回路卸压均会在堆芯熔化进程的3个阶段导致氢气产生率的明显增大：1）堆芯温度1500～2100K;2）堆芯温度2500～2800K;3）从形成由硬壳包容的熔融池（2800K）到熔融物向压力容器下封头下落。开启卸压阀的列数越多,氢气产生率的增大越明显。     

小型堆包壳破损状态下利用稳压器去除放射性研究

利用MELCOR程序对小型船用堆稳压器喷雾除气过程及停堆过程进行建模,进而模拟核动力装置从功率运行至降功率除气,以及除气结束后停堆消除稳压器气腔的全部物理过程。通过对反应堆关键运行参数变化趋势的仿真分析,验证了模拟的物理过程的合理性。结合建立的除气及停堆仿真模型,计算分析了包壳破损状态下,稳压器喷雾除气、停堆过程对稳压器内惰性气体含量的影响,评估了稳压器高点放气和喷雾除气对放射性物质的去除作用。研究结果能为小型堆包壳破损状态下放射性安全管理策略提供指导和帮助。     

瞬态条件下稳压器水位测量特性试验研究

稳压器是压水堆核动力装置压力安全系统的主要设备，其水位波动反映了一回路系统的水容积变化情况，是稳压器运行控制的关键参数之一。本文基于双区非平衡模型模拟蒸汽泄露条件下的稳压器水位变化，并针对稳压器蒸汽泄漏工况开展了水位测量特性试验研究，研究了2.6~7.8 kPa/s压降速率工况下，稳压器内水位测量压差的变化情况。研究发现:采用压差修正液相区密度计算的水位值在压力瞬变情况下有较好的跟随性，能够更好的反应水位特性；表征稳压器内液相区密度变化的压差在压力减小的过程中，过渡时间小于40 s，且过渡时间与压变速率单因素无强相关性。这为稳压器的安全运行控制提供了基础研究数据。      

PWR-1000XL核蒸汽供应系统设计

国产化1000MW级压水堆核电站（PWR－1000XL）是中国核动力研究设计院拟向国内用户推荐的计划在“十五”后期开始建造的核电站方案之一。PWR－1000XL的设计寿命60年，核蒸汽供应系统的主要设计特点是：采用Performanc^ 燃料组件，换料周期18个月：堆芯平均线功率密度165．2W/cm，堆芯热工裕量大于15％，堆顶结构一体化，设置RPV顶盖事故排气系统，无测温旁路系统；稳压器容积45m^3，选用△75型蒸汽发生器和100D型主泵；采用破前漏技术，设置可燃气体控制系统；采用数字化仪表和控制系统。     

PWR Cold-Leg Small-Break LOCA with Total HPI Failure

Cold-leg small-break loss-of-coolant accident (LOCA) tests were performed at the ROSA-IV Large Scale Test Facility (LSTF), a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). The tests were conducted for break areas ranging 0.5–10% of the scaled cold leg area, and simulated hypothetical total failure of the high pressure injection (HPI) system. One of the tests, conducted with 1% break area, included an intentional depressurization of the primary system that was initiated after the onset of core dryout. A simple prediction model is proposed for prediction of times of major events, namely, loop seal clearing, core dryout, accumulator (ACC) injection and actuation of low pressure injection (LPI) system. Test data and model calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of approximately 5% or more, and might be insufficient for intermediate break areas to maintain adequate core cooling. It is also shown that there might be possibility of core dryout after ACC injection and before LPI injection for break areas less than approximately 2.5%.     

Power-operated relief valve stuck-open accident and recovery scenarios in the Institute of Nuclear Energy Research integral system test facility

Four scaled small break loss-of-coolant accident (LOCA) tests simulating the pressurizer power-operated relief valves (PORVs) stuck-open accidents and the recovery actions in a pressurized water reactor (PWR) were performed at the Institute of Nuclear Energy Research (INER) integral system test (IIST) facility. The objectives of this study are to verify the effectiveness of emergency operating procedure (EOP) and emergency core cooling system (ECCS) on reactor safety. The break sizes were volumetrically scaled down based on one and all three fully-opened PORVs which is equivalent to 0.23% and 0.69% hot leg flow area of the reference plant. The experimental results indicate that in case of high pressure injection (HPI) system failure, the rapid depressurization of the steam generators is proved to be an effective way in the depressurization of the reactor coolant system and the core cooling. In contrast, if only one HPI charging pump operates normally, which injected half (or minimum) flow rate of normal cooling water, the core cooling can be adequately provided without operating the secondary bleeding during PORV stuck-open transient. This paper also presents the scaling methods for the reduced-height, reduced-pressure (RHRP) IIST facility and the test conditions. The validity of the present scaling methodology is confirmed by the results from previous IIST counterpart tests and comparison of the present results with those of the tests performed at the full-height, full-pressure(FHFP) stuck-open tests.     

大型非能动压水堆核电厂安全注射系统旋启式止回阀阻力变化对长期堆芯冷却影响分析

大型非能动压水堆核电厂在发生失水事故(LOCA)后的长期堆芯冷却阶段依靠重力向堆芯注入应急冷却水,其注射管线上设置的旋启式止回阀的阻力可随流量变化,管线的阻力可能将非预期地增加。根据旋启式止回阀阻力特性,为失水事故最佳估算系统分析程序添加相应的计算功能,对压力容器直接注射(DVI)管线双端断裂事故后长期堆芯冷却工况进行了计算分析。结果表明:安全注射管线上旋启式止回阀阻力变化对大型非能动压水堆核电厂LOCA后长期冷却的影响较小;在安全裕量不足的情况下,旋启式止回阀的阻力特性将影响到非能动注射管线的安全注射功能的执行。     

AP1000核电厂丧失正常给水稳压器防满水措施研究

在某AP1000核电厂丧失正常给水事件中，由于一系列的误操作导致稳压器满水，而稳压器安全阀在多次打开后可能无法重新关闭，不满足核电厂Ⅱ类工况验收准则。文章分析了该事件中稳压器满水的原因，即在非能动余热排出热交换器（PRHR HX）冷却能力充足的情况下，系统不适当的降压导致环路中冷却剂闪蒸，进而导致稳压器满水，此时通过开启堆顶放气阀启动应急下泄的方式无法有效降低稳压器液位。最后给出了AP1000核电厂丧失正常给水事故中防止稳压器满水的建议措施，即在RCS降压过程中，应确保RCS压力始终高于热管段温度对应的饱和压力，进而确保冷却剂不发生闪蒸。      

先进压水堆一回路地震反应谱分析及参数敏感性研究

压水堆一回路系统包含压力容器、蒸汽发生器、主泵、稳压器、主管道和波动管等重要部件,各部件在地震激励下的动态响应与整个系统的结构形式密切相关。本文从系统的角度,以非能动先进压水堆一回路为研究对象,运用ANSYS建立了其三维有限元模型,在模态分析的基础上,进行了三正交方向输入下的反应谱分析,得到了系统在地震载荷下的响应。并对反应谱输入角度和支撑刚度进行了敏感性研究,给出了这些特性参数对结构设计和分析的指导性意见。此外,通过直接积分法得到系统的地震时程响应,并与谱分析的模拟结果进行了对比,同时也为主泵等单个部件的详细地震分析提供位移、加速度输入。最后通过三维实体模型与集中质量模型抗震计算结果的比较,说明建立三维实体模型的必要性。本文为核电站一回路重要设备的结构分析提供了技术支持。     

AP1000核电厂反应堆冷却剂系统布置设计

AP1000核电厂反应堆冷却剂系统布置设计,在满足系统功能的前提下,充分考虑了屏蔽防护、核级部件在役检查、模块化设计、内部灾害防护等方面的要求。反应堆冷却剂系统主设备及主回路采用了紧凑型的布置方式,改善了环路配置的经济性,波动管布置在考虑足够柔性的基础上采用了大倾斜角连续上坡的方式,降低了波动管在运行过程中出现热分层的可能性,稳压器安全阀及ADS第1、2、3级集中布置在稳压器顶部,组合成一体化的模块Q601,改善了反应堆冷却剂系统布置结构。     

AP1000核电厂丧失正常给水稳压器防满水措施研究

In an accident of loss of feedwater in an AP1000 plant, the pressurizer was filled with water for a series of improper operations, and the safety valves may not be qualified to re-close following multiple cycles of opening, which is not acceptable in Condition Ⅱ events. The paper analyzes the causes for the filling of water in the pressurizer in this event, that is, the instantaneous evaporation of coolant in the loop during the process of improper depressurization of RCS while the PRHR HX is with sufficient cooling capability. At this time, the water level in the pressurizer level cannot be decreased by opening the reactor vessel head vent valves for emergency letdown. Finally, the recommended measure is provided to prevent the filling of water in the pressurizer during loss of normal feedwater for AP1000 NPP. The RCS pressure should always be higher than the saturation pressure corresponding to the temperature of the hot legs to avoid the coolant evaporation.     

全厂断电事故下AP1000非能动余热排出系统分析

利用RELAP5/MOD3.3程序对AP1000反应堆一回路及非能动系统进行建模计算,给出了AP1000非能动余热排出系统（PRHRS）在全厂断电事故下的瞬态响应特性。计算结果表明：情况1,PHRH系统由蒸汽发生器低水位与低启动给水流量符合信号启动,稳压器安全阀的开启导致PRHRS发生倒流现象,并会引起堆芯冷却剂过热沸腾、压力容器进出口温差过大等后果;情况2,由断电信号直接触发PRHRS,触发前安全阀不开启,此时PRHRS正常运行。