RELAP5/MOD3.2与MOD3.3临界流模型对比及分析

热工水力程序RELAP5/MOD3具有比较广泛的应用,文章基于RELAP5/MOD3.2与RELAP5/MOD3.3两个程序版本,对某反应堆冷段3.5in小破口失水事故进行计算分析,初步探讨不同临界流模型对计算结果的影响,相关结果可为分析类似小破口失水事故提供一定的参考。     

采用非能动余热排出系统实验数据对RELAP5程序的评价

利用非能动余热排出系统1∶10原理性实验台架的稳态实验与启动实验数据,对RELAP5/MOD3.2程序进行评估。结果表明：对于本原理性实验系统,RELAP5/MOD3.2程序过低估算了蒸汽流速对蒸汽凝结换热系数的影响,因而,程序中垂直管内的蒸汽凝结换热系数偏小,计算结果与实验结果偏差大。对RELAP5/MOD3.2程序垂直管内蒸汽凝结换热模型进行了修正,修正后的计算结果与实验值基本吻合。评价结果表明：采用RELAP5/MOD3.2程序对该类型的非能动余热排出系统进行计算,需对程序中垂直管内的蒸汽凝结换热模型进行修正。     

基于RELAP5的船用核动力装置二回路数字模型

根据核动力装置二回路系统一种新的结构设计方案,结合RELAP5/MOD3.4程序,建立了二回路汽轮机、冷凝器、给水泵及预热器等主要部件的物理模型。对二回路主要部件进行了单一部件模型适应性验证分析,并探讨了系统分析程序的局部计算能力。结果表明,RELAP5/MOD3.4程序模型的稳态计算结果与设计值基本吻合,动态计算也能够满足二回路主要部件的计算精度要求。     

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。     

PRS换热器蒸汽凝结换热模型研究

使用RELAP5/MOD3.2程序对某型核动力装置二次侧非能动余热排出系统(PRS)1:1实验装置进行稳态计算,一些工况下计算结果同实验结果偏差较大。研究了汽-液界面剪切应力及系统高压等条件对层流和湍流状态下竖直管内蒸汽凝结模型的影响,并对模型进行了改进。改进后的RELAP5程序对该系统1:1实验装置进行稳态和瞬态计算,计算结果同实验结果符合良好。     

非能动余热排出系统数学模型研究与运行特性分析

利用某型核动力装置非能动余热排出系统1:10原理性试验的8个稳态工况、6个启动工况的试验数据,验证RELAP5/MOD3.2程序对本类型非能动余热排出系统的适用性。结果表明:垂直管内蒸汽凝结换热系数对两相流自然循环的流动与传热影响大;RELAP5/MOD3.2程序过低估算了垂直管内蒸汽流速对蒸汽凝结换热系数的影响,计算结果与试验结果偏差大。对RELAP5/MOD3.2程序垂直管内的蒸汽凝结换热模型进行修正,修正后的计算结果与试验值基本吻合;采用RELAP5程序对垂直管内两相流自然循环传热进行计算,须选择热前沿跟踪模型。对非能动余热排出系统的稳态与瞬态运行特性进行分析,理论计算与试验结果均表明:稳态工况下,系统可以实现稳定的两相流自然循环,系统排热能力受蒸汽发生器水位的影响大,冷却水入口温度与系统压力的影响相对较小;系统的启动特性良好,可快速地建立环路的自然循环,带走反应堆的衰变热。     

RELAP5程序应用于二次侧非能动余热排出系统设计的初步评价

应用实验数据对RELAP5/MOD3.3程序应用于二次侧非能动余热排出系统设计进行初步评价,结果表明,系统投入初期,由于RELAP5程序的一维流动假设,针对事故冷却水池早期的内部三维对流传热模拟存在不足;系统投入后期,冷却水池内部传热为泡核沸腾,池式沸腾换热占主导地位,程序计算结果与实验结果较吻合,RELAP5/MOD3.3程序基本适用于二次侧非能动余热排出系统的稳态运行特性分析。     

RELAP5/MOD3.3程序对非能动核电厂小破口失水事故的适用性研究

AP1000核电厂采用非能动堆芯冷却系统来缓解小破口失水事故(SBLOCA),缓解事故的理念是流动冷却。RELAP5/MOD3.3程序适用于传统核电厂SBLOCA研究,对于非能动电厂SBLOCA研究的适用性需重新研究与评估。本工作基于非能动电厂小破口失水事故的分析,结合RELAP5/MOD3.3的结构与模型,对其进行评估和改进。为验证改进后的RELAP5/MOD3.3的适用性,以AP1000小破口失水事故的验证试验台架APEX-1000为模拟对象,分析模拟DBA-02、NRC-05事故工况。分析结果表明,改进后的RELAP5/MOD3.3的计算结果与试验数据符合较好。     

RELAP5／MOD2和CATHARE两相临界流模型的评价

分析和评价了2个程序中的两相临界流模型。指出RELAP5/MOD2在上游低欠热度或低含汽率条件下计算的临界流量偏低并不能反映几何尺寸(L/D)对临界流量的影响。CATHATE临界流模型较完善,它计算的临界流量与实验符合得很好。建议用非均匀热不平衡声速来修改RELAP5/MOD2的两相临界流判据,或者补充短喷管和孔板的临界流量实验关系式或实验数据,以改正RE-LAP5/MOD2的上述2个缺点。     

用RELAP5程序分析非能动堆芯应急冷却系统实验结果

先进压水堆的一个显著特点是非能动系统的高可靠性，评价这些系统的运行特性以及系统分析程序（如RELAP5等）的计算能力是非常重要的，中国核动力研究设计院设计建造了原理性的非能动堆芯应急冷却系统实验装置，并完成了相关实验研究，取得一批有价值的数据，本文用RELAP5／MOD3．2程序对实验过程进行了模拟分析。通过计算结果与实验结果的比较，初步评价了RELAP5／MOD3．2程序的计算能力。     

Control rod assembly drops to fully inserted position

This paper describes validation of a computer model that has been developed for VVER 440 Nuclear Power Plant (NPP) for use with RELAP5/MOD 3.2 computer code in the analysis of the following transient: “Control rod assembly drops to fully inserted position”.This validation is a process that compares the analytical results obtained by RELAP5/MOD3.2 model of the VVER 440 with experimental transient data received from Kozloduy NPP, Unit #2. The model of VVER 440 was developed at the the Institute for Nuclear Research and Nuclear Energy for analyses of operational occurrences, abnormal events and design basis scenarios. It will provide a significant analytical capability for the Bulgarian technical specialists located at the Kozloduy NPP.The criteria used in selecting transient are: importance to safety, availability and suitability of data followed by suitability for RELAP5 code validation. The comparisons between the RELAP5 results and the test data indicate good agreement.     

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.     

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).     

Evaluation of analytically scaled models of a pressurized water reactor using the RELAP5/MOD3 computer code

This paper performs analytical evaluations for the potential distortions caused by the scaled models using RELAP5/MOD3 computer codes. By use of scaling analysis, two scaled models with the same volumetric ratio are constructed for the Korean next generation reactor (KNGR), which is an advanced light water reactor. The scaling methodology adopted in this paper preserves the two-phase natural circulation similarities between prototype and scaled models. One scaled model is at full height with reduced flow area. The other model is at reduced height with reduced flow area. By using appropriate scale factors the RELAP5/MOD3 input models are developed. Then, the transient responses of the two ideal scaled models are simulated for small break loss of coolant accidents (SBLOCAs) by using the RELAP5/MOD3 computer code. The transient responses of the two scaled models are compared with those of the prototype. The results indicate that qualitative and quantitative similarities are well preserved for both models during SBLOCA with different break sizes.     

非能动专设安全设施应急余热排出试验研究

在海水淡化堆综合模拟试验装置上,开展了非能动专设安全设施应急余热排出模拟试验研究,获得了系统参数对非能动余热排出特性的影响规律。利用RELAP5/MOD3.2程序对蓄压水池不同初始水位下自然循环的建立和余热导出的过程进行了计算。结果表明,RELAP5/MOD3.2程序能较好地模拟海水淡化堆非能动专设安全设施的非能动余热导出过程,计算结果与试验结果符合较好。     

Development of a new bubbly-slug interfacial friction model for RELAP5

The development of a new bubbly-slug interfacial friction model for the Pressurized Water Reactor (PWR) safety code RELAP5 is described. The model is based on a set of best-estimate void fraction correlations which cover the full range of geometries and flow conditions encountered in PWR safety analysis. By exploiting the relationship between void fraction and interfacial friction that exists for steady, fully developed flow conditions, the correlations are converted into effective interfacial friction coefficients that can be applied in the code for transient as well as steady-state conditions. Assessments against separate effects tests indicate that the new model is more accurate than the existing model in many situations, particularly rod bundle geometries, and should never be significantly less accurate. The model has been implemented in a local version of RELAP5/MOD2 and in a pre-release version of RELAP5/MOD3 at Idaho National Engineering Laboratory (INEL).     

两流体模型在环形窄缝内的应用研究

采用EPRI最新开发的Chexal-Harrison相壁相间摩擦模型和简化的相壁相间传热模型,构造了适用于环形窄缝内沸腾传热和流动的两流体模型,并编制了热工水力计算程序——THYME程序.与实验数据比较,分析了环形窄缝套管在不同负荷下Relap5/Mod3.2程序和本文程序的计算结果.计算结果表明,Relap5/Mod3.2低估了环形蒸发管的蒸汽温度,本文计算结果与实验数据较为一致.     

蒸汽发生器完全丧失给水引发的压水堆严重事故研究

采用严重事故最佳估算程序SCDAP/RELAP5/MOD3.4,建立了美国Surry核电站的详细计算模型,对完全丧失给水(TLFW)引发的堆芯熔化事故进行了研究分析.为准确预测压力容器内堆芯熔化的进程,给二级PSA提供可信的初始条件,计算中考虑了一回路压力边界的蠕变破裂失效,并评价了人为干预对堆芯熔化进程及事故后果的影响.