Determination of BWR minimum steam cooling reactor water level and associated steam flow using MAAP4 code

The minimum steam cooling reactor pressure vessel (RPV) water level (MSCRWL) is defined to be the lowest RPV water level at which the covered portion of the core is capable of generating sufficient steam to preclude peak cladding temperature (PCT) in the uncovered portion of the core from exceeding 1500 °F. The associated steam flow rate is called Wg-1500. Both MSCRWL and Wg-1500 are important parameters for safe operation of nuclear power plant. In the past, the calculations of MSCRWL and Wg-1500 were calculated by indirect way via simple model and provided by the vendor. It has to be revised quite often during the operation period of nuclear power plant. The process is time-consuming. To improve the situation, a direct and easy method of generating MSCRWL using MAAP4 code by the utilities in conjunction with the proportional and integral (PI) controller is developed in this study. The developed control loop with a PI controller is capable of generating the MSCRWL in a fast and precise manner. The MSCRWL and Wg-1500 are calculated simultaneously by controlling the PCT equal to 1500 °F. Furthermore, the adjusting process is done automatically and readily with this methodology. The effect of feedwater inlet temperature is taken into account via thermal kits of the balance of plant system. The calculated MSCRWL is consistent with the calculated Wg-1500. The sensitivity study of power level and power shape are performed. For a given power shape, the MSCRWL is less sensitive to the power level. However, Wg-1500 is almost proportional to the power level. This information is helpful for the associated EOP application. This technique can be applied for other system codes.     

Modelling of Zry-4 cladding oxidation by air, under severe accident conditions using the MAAP4 code

In a nuclear power plant, a potential risk in some low probability situations in severe accidents is air ingress into the vessel. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of Fission Products (FP), especially increasing that of ruthenium. This FP is of particular importance because of its high radio-toxicity and its ability to form highly volatile oxides. Oxygen affinity is decreasing between Zircaloy cladding, fuel and ruthenium inclusions in the fuel. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues.A review of existing data in the field of Zircaloy-4 oxidation in air-containing atmosphere shows that this phenomenon is quantitatively well understood. The cladding oxidation process can be divided into two kinetic regimes separated by a breakaway transition. Before transition, a protective dense zirconia scale grows following a solid state diffusion-limited regime for which experimental data are well fitted by a parabolic time dependence. For a given thickness, which depends mainly on temperature and the extent of pre-oxidation in steam, the dense scale can potentially breakdown. In case of breakaway combined with oxygen starvation, cladding oxidation can then be much faster because of the combined action of oxygen and nitrogen through a complex self sustaining nitriding-oxidation process.A review of the pre-existing correlations used to simulate zirconia scale growth under air atmospheres shows a high degree of variation from parabolic to accelerated time dependence. Variations also exist in the choice of the breakaway parameter based on zirconia phase change or oxide thickness. Several correlations and breakaway parameters found in the literature were implemented in the MAAP4.07 Severe Accident code. They were assessed by simulation of the QUENCH-10 test, which is a semi-integral test designed to study fuel bundle exposure to steam first and then to air. This paper deals with the main results obtained with MAAP4.07 when simulating QUENCH-10.     

稳压器压力调节控制器动态特性的微机仿真

阐述了利用MATLAB软件的simulink仿真软件包实现对反应堆的稳压器压力调节控制器动态特性的微机仿真，得到了较为理想的仿真结果。     

稳压器水位控制改进研究

现代的反应堆稳压器的水位控制大都采用的是传统的PI调节器,对于这样一个非线性和时变性的复杂控制系统,PI调节器表现出来的常常是超调量大、调节时间长等并不理想的效果.本文提出以智能模糊控制器取代原有常规PI控制器的改进方案,并以大亚湾一期工程反应堆控制系统的稳压器水位控制为研究对象,着重阐述智能模糊控制原理在具体实践中的使用方法,最后对改造前后的系统输出进行LabVIEW仿真比较.仿真结果表明,模糊PI调节器不但克服了常规PI调节器的缺点,而且还提高了控制系统的实时性和抗干扰能力,值得广泛采用.     

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.     

核电站稳压器水位控制系统数字化的实现及优化方案

根据模拟稳压器水位控制系统结构、功能以及运行原理的分析,展开对稳压器水位控制系统数字化的研究。在选择成熟商用单板机的基础上引入了数/模,模/数转换以及控制算法等实现了稳压器水位控制系统,并和稳压器水位仿真模型构成闭环运行,通过虚拟仪器监测此控制系统的运行情况,证实了本系统达到并超过了原有模拟控制系统的功能和性能。     

Development of linked analysis system for PWR severe accidents utilizing RELAP5 and MAAP4

The purpose of this study is to develop a severe accident (SA) analysis method that is more reliable thorough transferring the physical status of the plant predicted by RELAP5 computer code to MAAP4 computer code. The methodology of the linkage analysis is developed and the criterion of linkage time is suggested to utilize the RELAP5 thermal–hydraulic calculation to the maximum degree possible and thereby guarantee the continuity of calculation for hydrogen generation. The MAAP4 calculations after data transfer show the physically proper results based on RELAP5 data. Comparison with other code results for TMI-2 accident reveals that the result from the RELAP5–MAAP4 linked analysis lay in the span given by a number of results of TMI calculation from other SA code systems. The results of this study are expected to improve the SA analysis methodology by analyzing an SA scenario with more reliable thermal–hydraulic initial conditions.     

基于acslX软件的AP1000稳压器压力控制系统动态仿真

针对AP1000稳压器内部传热传质过程的特点,结合西屋AP1000的相关参数,在上海核工程研究设计院控制系统模型的基础上,对其中稳压器的二区数学模型进行完善和改进,利用acslX软件建立稳压器三区动态数学模型,并严格按照西屋AP1000稳压器的压力控制逻辑,对建立起来的数学模型进行了相应的控制仿真实现。通过比较改进前和改进后模型试验结果与相关设计文件的差异,验证了改进后模型较改进前具有更好的精确性、可扩展性,同时该模型可为今后CAP1400稳压器的仿真工作打下一定基础。     

Photon production using a low energy electron expansion of the EGS4 code system

An expansion of the EGS4 code system has been developed in which electron-atom inelastic collisions are modelled taking into account atomic bound effects. Additional models have been developed to deal with the ionisation of the K-shell and improved angular sampling of newly created bremsstrahlung photons. The incorporation of these features in EGS4 is described and the updated code is used to simulate photon generation in X-ray tubes. The accuracy of the new code and some of its possible applications are discussed.     

Application of artificial intelligence techniques in modeling and control of a nuclear power plant pressurizer system

In pressurized water reactor (PWR) nuclear power plants (NPPs) pressure control in the primary loops is fundamental for keeping the reactor in a safety condition and improve the generation process efficiency. The main component responsible for this task is the pressurizer. The pressurizer pressure control system (PPCS) utilizes heaters and spray valves to maintain the pressure within an operating band during steady state conditions, and limits the pressure changes during transient conditions. Relief and safety valves provide overpressure protection for the reactor coolant system (RCS) to ensure system integrity. Various protective reactor trips are generated if the system parameters exceed safe bounds. Historically, a proportional-integral-derivative (PID) controller is used in PWRs to keep the pressure in the set point, during those operation conditions. The purpose of this study is two-fold: first, to develop a pressurizer model based on artificial neural networks (ANNs); secondly, to develop fuzzy controllers for the PWR pressurizer modeled by the ANN and compare their performance with conventional ones. Data from a 2785 MWth Westinghouse 3-loop PWR simulator was used to test both the pressurizer ANN model and the fuzzy controllers. The simulation results show that the pressurizer ANN model responses agree reasonably well with those of the simulated power plant pressurizer, and that the fuzzy controllers have better performance compared with conventional ones.     

Investigating the execution of EOPs in LOTDFP + ATWS + LOCA scenarios for an ABWR using MAAP code

The effectiveness of the execution of emergency operation procedures (EOPs) for an advanced boiling water reactor (ABWR) during postulated accident conditions using MAAP 4 code is discussed in this paper. The simulation scenarios included the loss of turbine driven feedwater pump (LOTDRFP), the anticipated transients without scram (ATWS), and the loss of coolant accident (LOCA). Based on the comparisons of responses on different parameters for cases with and without EOP actions, we concluded that the EOPs could effectively mitigate the consequences of the accidents. In addition, the emergency depressurization (ED) timing and the times spent between executing the EOP steps were also considered. The simulation results clearly reveal that both the earlier execution of ED and the decrease of times spent between each EOP step could delay the boron injection and leave the operator ample time to take some other remedy actions for reactor safe shutdown.     

SSRF高频低电平系统预制研究

上海同步辐射装置(SSRF)预研阶段完成了一套500MHz高频系统的低电平控制系统,它主要由信号源、速调管的激励控制、三个反馈控制环路和连锁保护构成。本文重点介绍三个反馈环路的设计、测量和调试结果。     

基于MAAP5程序的秦山核电站严重事故分析

应用MAAP5程序建立了秦山核电站一、二回路,安全系统以及安全壳的模型,并以冷段双端断裂叠加高高、高、低压安注失效,安全壳喷淋系统失效为例,对该严重事故序列进行了模拟计算,给出了瞬态过程一些重要参数随时间的变化规律。结果表明：在72 h内无能动干预手段的条件下,安全壳的完整性可得到保证,相关数据可为秦山核电站严重事故预防和事故缓解措施的制定提供重要参考。     

Calculation of the void reactivity of CANDU lattices using the code system

The reactivity effect of coolant voiding in CANDU-type fuel lattices has been calculated with different methods using the code system. The known positive void reactivity coefficient of the original lattice was correctly obtained. A modified fuel bundle containing dysprosium and slightly enriched uranium to eliminate the positive reactivity effect was also calculated. Owing to the increased heterogeneity of this modified fuel the one-dimensional cylindrical calculation with XSDRN proved to be inadequate. Code options allowing bundle geometry were successfully used for the calculation of the strongly space dependent flux and spectrum changes which determine the void reactivity.     

Blowdown force analysis of piping system under LOCA conditions using BLOWDOWN code

The BLOWDOWN code was developed for blowdown force analysis of piping system under LOCA conditions. This is a post-processor of the thermal-hydraulic analysis code RELAP4/MOD6. The results obtained from the RELAP4/MOD6 code are converted into blowdown forces by the BLOWDOWN code.In the paper, the physics and algorithms of the BLOWDOWN code are outlined. Some numerical examples are also presented to show the effectiveness of the code.     

Analysis of LMFBR primary system response to an HCDA using a eulerian computer code

This paper describes the use of the Eulerian code to predict the response of the fast reactor containment and the primary piping loops to HCDAs, and to analyze the sodium-spillage and the bubble motion problems which cannot be analyzed with a Lagrangian code. The basic equations and numerical techniques used in the Eulerian computer code are described in detail. Four sample problems are given. The first problems are given. The first problem deals with the response of the primary containment to an HCDA. The second problem deals with the slug impact on the reactor head and the spillage of sodium through the opening holes. The third problem deals with the dynamics of an HCDA bubble. The fourth problem deals with the response of a piping loop.     

Analysis of the primary containment response using a hydrodynamic-elastic-plastic computer code

The dynamic response of the primary reactor containment system to a hypothetical core disruptive accident (HCDA) is determined from the basic equations of mass, momentum, and energy, and the equations of state of the medium. These equations are first expressed in material coordinates and then set into finite difference form solved numerically on the computer using a hydrodynamic-elastic-plastic computer code, REXCO-HEP developed at ANL. Propagation of pressure waves, loads imposed on different parts of the reactor components, and the resulting deformations are determined at every time step throughout the sequence of the calculation. As a sample calculation, the code was applied to analyze the response of the FFTF reactor to a 150 MWsec HCDA. The mathematical model is described in detail, particularly in the areas of modeling reactor internals and extending the time range to cover the entire excursion phenomenon. Finally, the results obtained from the computer analysis are discussed in detail.     

Transient analyses for accelerator driven system PDS-XADS using the extended SIMMER-III code

Transient analyses for Preliminary Design Studies of an Experimental Accelerator Driven System (PDS-XADS) were performed with the reactor safety analysis code SIMMER-III, which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors so as to describe the XADS specifics such as subcritical core, strong external neutron source and lead–bismuth–eutectic (LBE) coolant. As transient scenarios, the following cases were analyzed in accordance with the PDS-XADS program: spurious beam trip (BT), unprotected beam overpower (UBOP), unprotected transient overpower (UTOP), unprotected loss of flow (ULOF) and unprotected blockage (UBL) in a single fuel assembly. In addition, to cover some core-melt situations and investigate the potential for recriticalities, so-called snap-shot analyses with ad hoc postulated severe blockage conditions were also investigated.The simulation results for BT and UBOP showed that immediate fuel damage might not take place under short-time beam interruption or a 100% increase of the external neutron source. Concerning UTOP, it was found that a reactivity jump of 1 $ would not lead to damage of the fuel and the cladding. The ULOF simulation showed that the remaining natural convection of the coolant would prevent the cladding from disruptions. In the simulation of UBL in a single fuel assembly, it was shown that no cladding failure might be expected, due to the radial heat transfer and the coolant flow in the hexcan gap. Under an artificial suppression of the radial heat transfer for this UBL case, a pin failure occurred in the simulation but subsequent fuel sweep-out into the upper plenum region would bring a reactivity reduction and no power excursion. The severe accident simulations starting from postulated blockage above an already disrupted core showed that a severe recriticality could be avoided by the fuel sweep-out into the dummy-assembly or hexcan gap regions.The present simulation results showed that the current PDS-XADS design has a remarkable resistance against severe transient scenarios even in core-degradation conditions.     

Gamma-ray shielding study of light to heavyweight concretes using MCNP-4C code

In this work, linear and mass attenuation coefficients, half and tenth-value layers, effective atomic number and electron density of different types of concretes were determined at 316.51, 468.07, 511, 662, 1173 and 1332 keV using MCNP-4C code and Win XCom programs. The MCNP-4C and Win XCom results agreed well with each other, with differences of ±1.9%. The results agreed with available experimental data, too, with differences of ±6%.The MCNP-4C results showed better agreement with the experimental data than the Win XCom results. Also, it was found that the effective electron density of studied concretes varies in the range of(2.83–3.2) 9 10~(23)electron/g.