核电厂运行瞬态分析软件PANTO的研发

针对核电厂运行瞬态分析的功能需求,中国核动力研究设计院研发了PANTO(Program for Analysis of Normal Transient and Overpressure)软件。该软件基于成熟可靠的系统分析模型和特殊部件模型,采用模块化的软件设计理念,应用面向对象的C++语言和java语言,具有完全自主知识产权。PANTO软件通过了单元测试、集成测试和系统测试,基本消除了所有的代码缺陷。针对秦山二期核电厂阶跃负荷增大10%与额定功率下全部甩负荷瞬态试验进行了验证计算。结果表明,PANTO软件能够较好地模拟瞬态中关键参数的变化情况,计算精度满足工程应用要求,适用于压水堆核电厂运行瞬态分析。     

Comparison of RELAP5 calculations of VVER-1000 coolant transient benchmark phase 1 at different power

This paper provides comparisons between experimental data of “MCP switching on when the other three MCPs are in operation” and RELAP5 calculations with different initial levels of the reactor power 29.45% and 27.47% from the nominal.

The reference power plant for this analysis is Unit 6 at the Kozloduy nuclear power plant (NPP) site. RELAP5/MOD3.2 computer code has been used to simulate the investigated transient. Operational data from Kozloduy NPP have been used for the purpose of assessing how the RELAP5 model compares against plant data. During the plant-commissioning phase at Kozloduy NPP Unit 6 a number of experiments have been performed. One of them is switching on MCP when the other three MCPs are in operation.

This investigation was possible through the participation of leading specialists from Kozloduy NPP and with the support of Argonne National Laboratory, under the International Nuclear Safety Program (INSP) of the United States Department of Energy.     

Simulation of the pressure set-point change transient with the Kuosheng plant analyzer

The purpose of this paper is to simulate the pressure set-point change transient with the Kuosheng plant analyzer. The Kuosheng plant analyzer was developed for the Kuosheng nuclear power station based on the boiling water reactor plant analyzer of the Brookhaven National Laboratory and the Chinshan plant analyzer of the Taiwan Power Co. First, the pressure control system (PCS) model was verified on a stand-alone basis, with the sensed dome pressure test signal as the input variable. Then, a transient induced by a step change in the pressure set-point was simulated with the Kuosheng plant analyzer and compared with the test data. The predicted results of the main system variables showed good agreement with the test data-within 8% of the maximum relative deviation. In addition, the PCS performance of the Kuosheng nuclear power station was analyzed.     

Analytical one-dimensional frequency response and stability model for PWR nuclear power plants

A dynamic model for PWR nuclear power plants is presented. The plant is assumed to consist of a one-dimensional single-channel core, a counterflow once-through steam generator (represented by two nodes according to the non-boiling and boiling region) and the necessary connecting coolant lines. The model describes analytically the frequency response behaviour of important parameters of such a plant with respect to perturbations in reactivity, subcooling or mass flow (both at the entrances to the reactor core and/or the secondary steam generator side), and perturbations in steam load or system pressure (on the secondary side of the steam generator). From corresponding ‘open’ loop considerations, it can then be concluded - by applying the Nyquist criterion - upon the degree of the stability behaviour of the underlying system. Based on this theoretical model, a computer code named ADYPMO has been established.From the knowledge of the frequency response behaviour of such a system, the corresponding transient behaviour with respect to a stepwise or any other perturbation signal can also be calculated by applying an appropriate retransformation method, e.g. by using the digital code FRETI. To demonstrate this procedure, a transient experimental curve measured during the pre-operational test period at the PWR nuclear power plant KKS Stade was recalculated using the combination ADYPMO-FRETI. Good agreement between theoretical calculations and experimental results give an insight into the validity and efficiency of the underlying theoretical model and the applied retransformation method.     

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.     

CPR1000全厂断电事故瞬态特性分析

用RELAP5/MOD3.4程序对CPR1000压水堆一回路系统进行整体建模,分析全厂断电事故下一回路主要参数的瞬态热工水力特性,并将RELAP5模型计算结果与THEMIS程序的计算结果进行对比,二者符合得较好。计算结果表明：该模型可较准确地模拟CPR1000在事故下的热工水力特性。     

福岛条件下AP1000核电厂事故前期研究

假设AP1000核电厂发生类似福岛核事故的初因事件,利用RELAP5/MOD3.3程序对事故早期的一、二回路系统和非能动安全系统进行模拟计算,得到了反应堆冷却剂系统压力、堆芯冷却剂温度、非能动安全系统流量等重要参数的瞬态变化。分析表明：在非能动余热排出系统完好的情况下,反应堆系统能顺利进入热停堆状态;如果非能动余热排出系统1根换热管发生双端断裂,则反应堆系统将会在5 h内发生严重事故。     

AP1000核电厂主给水管道断裂事故瞬态特性分析

AP1000是目前国际上典型的“三代”非能动核电厂,基于最佳估算程序RELAP5/MOD3.3,对AP1000核电厂系统进行了详细的建模分析,获得了主给水管道断裂事故下AP1000核电厂关键参数的瞬态特性和非能动系统响应特性。结果表明,事故过程中一、二回路的压力和温度呈现波动变化,一回路压力最大值为17.13 MPa,低于设计压力的91%,主蒸汽系统的压力也低于设计值的91%,满足验收准则的要求。     

Monte Carlo simulation of molecular flow in a neutral beam injector and comparison with experiment

Monte Carlo calculations have been performed to obtain estimates of the background gas pressure and molecular number density as a function of position in the PDX-prototype neutral beam injector, which has undergone testing at the Oak Ridge National Laboratory. Estimates of these quantities together with the transient and steady-state energy deposition and molecular capture rates on the cryopanels of the cryocondensation pumps and the molecular escape rate from the injector were obtained utilizing a detailed geometric model of the neutral beam injector. The molecular flow calculations were performed using an existing Monte Carlo radiation transport code, which was modified slightly to monitor the energy of the background gas molecules. The credibility of these calculations is demonstrated by the excellent agreement between the calculated and experimentally measured background gas pressure in front of the beamline calorimeter located in the downstream drift region of the injector. The usefulness of the calculational method as a design tool is illustrated by a comparison of the integrated beamline molecular density over the drift region of the injector for three modes of cryopump operation.     

AP1000核电厂自动泄压系统误启动事故计算

基于最佳估算程序RELAP5/MOD3.3,对AP1000核电厂冷却剂系统和非能动堆芯冷却系统进行了建模分析,得到了自动泄压系统(ADS)阀门误启动事故下,系统压力、破口流量、系统水装量等参数的瞬态变化,计算结果与西屋公司采用NOTRUMP程序的计算结果进行了比较与分析。结果表明:AP1000核电厂的非能动专设安全设施能有效对一回路进行冷却和降压,防止堆芯过热,验证了AP1000发生ADS阀门误启动事故后的安全性。     

Modelling of a supercritical CO2 power cycle for nuclear fusion reactors using RELAP5–3D

A supercritical recompression CO₂ power cycle has been simulated using the system code RELAP5–3D. This code is being developed by INL and has traditionally been used in the simulation of operational and accidental transients in fission nuclear plants. The aim of the work presented here, developed within the framework of the Spanish Fusion Technology Program Consolider TECNO_FUS, is to take advantage of the simulation capabilities of RELAP5–3D in a field where little if any experience exists in the use of the code; i.e., the simulation of the heat fluxes and the thermodynamic cycle that, in a fusion power plant, will convert thermal power from plasma into mechanical power as a previous step to electricity generation. Code capabilities that make it suitable for this purpose are, for instance, the compressor model and the libraries of fluid properties (among them CO₂ and LiPb).The reference plant for the simulation is the one being designed under TECNO_FUS, which is the Spanish proposal for DEMO. The model of the plant includes the primary coolant systems, i.e. helium and LiPb in the Spanish dual coolant modular design (doble refrigerante modular, DRM), compressors, turbine and heat exchangers (Printed Circuit type).After the model has been set-up, several steady-state calculations have been run to test the performance of the model. After designing some minimal control features and adjusting their parameters, a few transient calculations have been run in order to demonstrate the capabilities of code and model. Finally, strengths and weaknesses of code and model are highlighted, along with some conclusions on their suitability for fusion technology applications.     

先进堆非能动余热排出系统应对全厂断电事故的能力分析

采用RELAP5/MOD程序对先进堆全厂断电事故进行分析计算,论证非能动余热排出系统对事故的缓解能力.分析表明,先进堆在发生全厂断电事故后,完全能够依靠非能动余热排出系统导出堆芯余热,保证反应堆的安全;先进堆非能动余热排出系统的设计总体上是成功的.     

直接布雷顿循环气冷反应堆系统运行特性分析

基于Matlab/simulink程序，针对小型直接布雷顿循环反应堆系统，通过模块化思想建立该系统数学物理模型，开发了系统分析程序。通过改变反应堆、透平、压缩机、换热器等关键设备的运行参数或引入阶跃扰动，模拟了系统稳态工况与瞬态变工况运行，得到了关键设备功率、进出口压力、温度等关键参数的变化曲线。结果表明，系统分析程序对小型直接布雷顿循环反应堆系统稳态与瞬态运行特性的模拟结果较合理，能为小型直接布雷顿循环反应堆系统的设计、优化与安全分析提供依据。     

Sensitivity analysis of the MASLWR helical coil steam generator using TRACE

Accurate simulation of transient system behavior of a nuclear power plant is the goal of systems code calculations, and the evaluation of a code's calculation accuracy is accomplished by assessment and validation against appropriate system data. These system data may be developed either from a running system prototype or from a scaled model test facility, and characterize the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs. The validation and assessment of the best estimate thermal hydraulic system code TRACE against the Multi-Application Small Light-Water Reactor (MASLWR) Natural Circulation (NC), helical coil Steam Generator (SG), Nuclear Steam Supply System (NSSS) design is a novel effort, and is the topic of the present paper. Specifically, the current work relates to the assessment and validation process of TRACE code against the NC database developed in the OSU-MASLWR test facility. This facility was constructed at Oregon State University under a U.S. Department of Energy grant in order to examine the NC phenomena of importance to the MASLWR reactor design, which includes an integrated helical coil SG. Test series have been conducted at this facility in order to assess the behavior of the MASLWR concept in both normal and transient operation and to assess the passive safety systems under transient conditions. In particular the test OSU-MASLWR-002 investigated the primary system flow rates and secondary side steam superheat, used to control the facility, for a variety of core power levels and Feed Water (FW) flow rates. This paper illustrates a preliminary analysis, performed by TRACE code, aiming at the evaluation of the code capability in predicting NC phenomena and heat exchange from primary to secondary side by helical SG in superheated condition and to evaluate the fidelity of various methods to model the OSU-MASLWR SG in TRACE. The analyses of the calculated data show that the phenomena of interest of the OSU-MASLWR-002 test are predicted by the code and that one of the reasons of the instability of the superheat condition of the fluid at the outlet of the SG is the equivalent SG model used to simulate the different group of helical coils. The SNAP animation model capability is used to show a direct visualization of selected calculated data.     

TRAC-PF1/MOD1 calculations and data comparisons for mist small-break LOCA, feed-and-bleed and steam-generator tube-rupture experiments

Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents, loss of feedwater and other transients in Babcock and Wilcox (B&W) nuclear plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 × 4 [two hot legs and steam generators (SGs), four cold legs and reactor coolant pumps] representation of B&W lowered-loop reactor systems. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Efforts are under way at Los Alamos to assess TRAC-PF1/MOD1 against data from the MIST facility.Calculations and data comparisons for TRAC-PF1/MOD1 assessment are presented for three transients run in the MIST facility. The energy removal and depressurization mechanisms in these tests are identified and the phenomena occurring in these tests compared. The tests analyzed are MIST Test 3109AA, the nominal small-break LOCA, Test 330302, a feed and bleed test with delayed high-pressure injection; and Test 3404AA, an SG tube-rupture test with the affected SG isolated. TRAC was able to predict these phenomena although the timing and magnitude of events were not always in good agreement.The MIST test have demonstrated the thermal-hydraulic phenomena expected to occur during transients in B&W nuclear plants. Because of scaling atypicalities, test results cannot be extrapolated directly to plant conditions. Although the phenomena were demonstrated in the MIST tests, there may be differences in the timing, magnitude and sequences of events in plant transients. Assessment calculations, three of which are presented here, have shown that the TRAC computer code can predict the major trends and phenomena occurring during the MIST tests with reasonable qualitative agreement. This includes complex sequences of events. Reasonable qualitative agreement is defined as meaning that major trends are predicted correctly, although TRAC values are frequently outside the range of data uncertainty. These assessment results, taken with assessment results from other facilities at a wide range of scales, provide us with confidence that the TRAC code can adequately simulate the transient phenomena possible in B&W nuclear plants.     

Computational investigation of the temperature and geometric characteristics of VVÉR fuel in the Kozlodui nuclear power plant (Bulgaria) under normal operating conditions

A method of performing stationary thermomechanical calculations of VVéR-440 and-1000 fuel elements, using the TRANSURANUS computer code to obtain the dependence of the temperature and radius of the fuel elements on the lineal power ensity and burnup, is described. These dependences are intended for use in neutron-physical calculations of the VVéR reactor at the Kozlodui nuclear power plant in stationary and transient regimes. The results obtained with this computer program are compared with calculations performed using the certified TOPRA-s code. The comparison shows reasonable agreement between the results of calculations of the fuel temperature. __________ Translated from Atomnaya énergiya, Vol. 101, No. 5, pp. 336–342, November, 2006.     

Advanced treatment of zircaloy cladding high-temperature oxidation in severe accident code calculations: PART II. Best-fitted parabolic correlations

Pressurized water vessel-type reactor (VVER) safety has become a very important issue, in particular for countries in Central and Eastern Europe. For thermal-hydraulic analyses the western codes like RELAP5, CATHARE and ATHLET were used.The purpose of the study was to quantitatively assess the RELAP5 capability to predict the main circulation pump (MCP) trip at nearly full power transient in Mochovce VVER 440/213 nuclear power plant (NPP). The transient parameters were recorded during the start up test program implementation. For accuracy quantification the improved fast Fourier transform based method (FFTBM) was used. The RELAP5/MOD3.2.2 computer code was used for calculation. The results showed very good agreement between calculated and plant measured data. The results also confirmed some previous studies that the simpler is the transient the higher code accuracy is generally achieved.     

RELAP5/MOD3.2 sensitivity calculations of loss-of-feed water (LOFW) transient at Unit 6 of Kozloduy NPP

This paper provides a comparison between the real plant data obtained by Unit 6 of Kozloduy nuclear power plant (NPP) during the loss-of-feed water (LOFW) transient and the calculation results received by RELAP5/MOD3.2 computer model of the same NPP unit.RELAP5/MOD3.2 computer model of the VVER-1000 has been developed at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) based on Unit 6 of Kozloduy NPP. This model has been used for simulation the behavior of the real VVER-1000 NPP during the LOFW transient. Several calculations have been provided to describe how the different boundary conditions reflect on the prediction of real plant parameters.This paper discusses the results of the thermal–hydraulic sensitivity calculations of loss-of-feed water transient for VVER-1000 reactor design. The report also contains a brief summary of the main NPP systems included in the RELAP5 VVER model and the LOFW transient sequences.This report was possible through the participation of leading specialists from Kozloduy NPP and with the assistance of Argonne National Laboratory (ANL) for the United States Department of Energy (US DOE), International Nuclear Safety Program (INSP).     

Animation model of Krško nuclear power plant for RELAP5 calculations

Today most software applications, also in the nuclear field, come with a graphical user interface. The first graphical user interface for the RELAP5 thermal-hydraulic computer code was called the Nuclear Plant Analyzer (NPA). Later, Symbolic Nuclear Analysis Package (SNAP) was developed. The purpose of the present study was to develop SNAP animation model of Krško nuclear power plant (NPP) for RELAP5 calculations with the aim to help analyze the results. In addition, the reference calculations for Krško full scope simulator validation were performed with the latest RELAP5/MOD3.3 Patch 03 code and compared to previous RELAP5 versions to provide verified source data, needed to demonstrate animation model. In total six scenarios were analyzed: two scenarios of the small-break loss-of-coolant accident, two scenarios of the loss of main feedwater, a scenario of the anticipated transient without scram, and a scenario of the steam generator tube rupture. The use of SNAP for animation of Krško nuclear power plant analyses showed several benefits, especially better understanding of the calculated physical phenomena and processes. It can be concluded that an animation tool was created, which enables to analyze very complex accident scenarios. The graphical surface helps keeping the overview and focusing on the main influences. Also, the use of such support tools to system codes may significantly contribute to better quality of safety analysis.