LOFTL2-5LBLOCA试验分析方法

采用保守评价模型与电厂状态参数最佳估算相结合的方法对大破口冷却剂丧失事故( LBLOCA)进行认证分析.以RELAP5/MOD3为分析工具,结合非参数统计方法,对电厂状态参数进行不确定性量化分析,对LOFT L2-5冷段双端剪切断裂LBLOCA整体试验进行了冷却剂丧失事故(LOCA)分析.分析表明,引入保守分析模式与最...     

风险指引的CPR1000核电厂LBLOCA分析方法初步研究

随着核电厂安全分析方法的不断发展，结合传统确定论分析与概率风险评价（PSA）的风险指引型安全分析方法逐渐引起安审当局和核电业主的广泛关注。本文基于国际上风险指引型分析方法在其他领域的应用现状，提出了风险指引的大破口失水事故（LBLOCA）分析方法，并重新评估了CPR1000核电厂的堆芯燃料包壳峰值温度（PCT）裕量。在PSA分析中，识别并量化了LBLOCA发生后可能发生的162个事件序列，并采用确定论现实分析方法（DRM）对筛选出的18个概率较大的事件序列进行了计算分析。然后通过期望值评估法和特定序列覆盖法对LBLOCA的PCT裕量进行了评估。结果表明，本文方法下LBLOCA的PCT裕量约为36~55 ℃，相比于传统的DRM裕量提升了16~35 ℃。     

现实LOCA分析中不确定性量化分析方法研究

基于中国改进型三环路压水堆(CPR1000)核电厂最佳估算热工水力系统分析程序(CATHARE GB)大破口失水事故(LBLOCA)分析模型,采用多种不确定性量化分析方法量化LBLOCA分析结果的不确定性,并针对不同方法对输入参数的处理、分析过程和输出结果的处理3个方面进行比较。结果表明,敏感性分析方法由于存在人为的保守性,所以其计算结果最保守,传统的参数统计方法、欧文因子法和Bootstrap方法需进行输出结果的正态性检验,而Wilks方法不需要进行正态性检验且分析结果更加现实。     

大破口失水事故分析方法GSM的开发与应用

中广核确定论统计方法(GSM)是介于保守评价模型和最佳估算评价模型之间的失水事故(LOCA)分析方法。在该方法中,程序模型采用确定论现实方法(DRM)惩罚模型进行保守方法处理,对电厂模型采用保守假设,对电厂重要状态参数采用统计方法量化确定不确定性范围和分布,并对统计抽样计算得到的目标参数分别采用参数统计和非参数统计处理以得到包壳峰值温度的双95%值上限值。将该方法应用于CPR1000核电厂大破口LOCA分析,与传统DRM相比可挖掘约9%的LOCA裕量。     

失水事故分析程序临界流模型改进及验证

失水事故（LOCA）分析中保守分析方法不利于提高核电厂的经济性，为了满足10CFR50附录K的核电厂LOCA评价要求，基于最佳估算程序RELAP5对其模型进行修改以满足对LOCA的评价要求，同时增大设计裕量。由于附录K涉及模型较多，本文主要对LOCA模型修改和验证方法进行研究，改进了RELAP5程序临界流模型，添加保守的Moody两相临界流模型，同时增加过冷临界流Zaloudek模型，并分别采用分离效应实验装置Marviken、Edward喷放管和整体效应装置Bethsy对程序进行了验证，结果表明添加的模型对模拟喷放过程临界流现象具有足够的可靠性。      

COSINE系统分析程序模型评估需求分析

COSINE是我国首个完全自主开发的用于核反应堆设计与安全分析的软件包,其系统分析程序具有保守模型与最佳估算模型两个版本。依据国际最新的评价模型开发与评估方法——EMDAP方法,对COSINE系统程序的保守模型和应用于最佳估算大破口失水事故(LOCA)事故分析的最佳估算模型所需评估的重要现象和过程进行识别和排序,制定出大破口LOCA事故PIRT表。同时,根据模型评估需求,构建核电软件模型评估数据库。     

大破口失水事故分析方法GSM的开发与应用

中广核确定论统计方法（GSM）是介于保守评价模型和最佳估算评价模型之间的失水事故（LOCA）分析方法。在该方法中，程序模型采用确定论现实方法（DRM）惩罚模型进行保守方法处理，对电厂模型采用保守假设，对电厂重要状态参数采用统计方法量化确定不确定性范围和分布，并对统计抽样计算得到的目标参数分别采用参数统计和非参数统计处理以得到包壳峰值温度的双95%值上限值。将该方法应用于CPR1000核电厂大破口LOCA分析，与传统DRM相比可挖掘约9%的LOCA裕量。     

大亚湾核电站18个月换料大破口失水事故的计算分析

大亚湾核电站18个月换料的设计中，堆芯焓升因子和功率峰值因子有了较大的提高，通过采用DRM分析方法和CATHARE程序对LBLOCA事故进行了较为全面的计算、分析和论证，得出了在18个月换料运行方式下，堆芯的包壳温度等参数仍然满足验收准则的结论。在此基础上重新建立了LOCA包络限制线。     

快堆假想堆芯解体事故程序研发

钠冷快中子反应堆是以钠作为冷却剂的第4代核能系统之一,为保证快堆在严重事故下能够包容放射性物质,对快堆假想堆芯解体事故进行准确模拟计算是非常必要和迫切的。采用改进型B-T模型对快堆假想堆芯解体事故进行分析是目前国际上主要的分析方法,为能更好地分析快堆假想堆芯解体事故,在改进型B-T模型的基础上引入快堆实际的堆芯反应性系数分布函数。本工作与法国的EPIXCOPOS程序计算结果的对比验证表明,程序模型能对快堆假想堆芯解体事故进行保守分析。     

压水堆核电厂失水事故后安全壳内产氢量计算研究

采用ORIGEN2程序对压水堆核电厂失水事故工况下堆芯区和地坑区氢气的产生量进行计算,以合理减少安全壳内可燃气体的控制设计评价的保守性.通过冷却剂的辐照分解产氢以及其他相关计算模型,对600MW(电功率)级压水堆核电厂失水事故工况下的氢气产生量进行计算.计算结果表明原评价结果过于保守,在核电厂失水事故后仍有充分的时间准备投入安全壳内氢气复合器.     

Development and application of a deterministic-realistic hybrid methodology for LOCA licensing analysis

It is well recognized that a realistic LOCA analysis with uncertainty quantification can generate greater safety margin as compared with classical conservative LOCA analysis using Appendix K evaluation models. The associated margin can be more than 200 K. To quantify uncertainty in BELOCA analysis, generally there are two kinds of uncertainties required to be identified and quantified, which involve model uncertainties and plant status uncertainties. Particularly, it will take huge effort to systematically quantify individual model uncertainty of a best estimate LOCA code, such as RELAP5 and TRAC. Instead of applying a full ranged BELOCA methodology to cover both model and plant status uncertainties, a deterministic-realistic hybrid methodology (DRHM) was developed to support LOCA licensing analysis. Regarding the DRHM methodology, Appendix K deterministic evaluation models are adopted to ensure model conservatism, while CSAU methodology is applied to quantify the effect of plant status uncertainty on PCT calculation. Generally, DRHM methodology can generate about 80-100 K margin on PCT as compared to Appendix K bounding state LOCA analysis.     

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.     

Boiling Water Reactor Loss-of-Coolant Tests with Two Bundle Loop

Two simulation tests for a boiling water reactor large loss-of-coolant accident (LOCA), conducted in the two bundle loop, were analyzed using the current licensing code system. These tests were recirculation-pump suction-line double-ended break tests. One of these tests assumed failures for LPCS and 1 out of 3 LPCIs, and another test assumed HPCS failure. A main objective of these analyses is to confirm the conservativeness of the licensing analysis models. Conclusions reached from the analyses are as follows:

1. Calculated heater surface temperature begins to rise much earlier than the measured temperature, due to the conservative GEXL model for a LOCA analysis.

2. Calculated heat up rate is higher than the test data, mainly due to neglecting the steam cooling in the analysis.

3. Calculated heater rewetting time is later than the test data, due to neglect of counter current flow limiting at the core inlet, when the measured ECCS flow rate is used in the analysis.

It has been confirmed that the current licensing analysis models give a conservative result for peak cladding temperature (PCT), due to the model conservativeness factors presented above, when the measured data are used in the analysis for the outflow from the system and the inflow to the system.     

Development of LOCA licensing calculation capability with RELAP5-3D in accordance with Appendix K of 10 CFR 50.46

In light water reactors, particularly the pressurized water reactor (PWR), the severity of a loss of coolant accident (LOCA) will limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the PCT evaluation during LOCA, it generally takes more resources to develop. Instead, implementation of evaluation models required by Appendix K of 10 CFR 50 upon an advanced thermal–hydraulic platform also can gain significant margin for the PCT calculation. The compliance of the current RELAP5-3D code with Appendix K of 10 CFR 50 has been evaluated, and it was found that there are ten areas where code assessment and/or further modifications were required to satisfy the requirements set forth in Appendix K of 10 CFR 50. All of the ten areas have been further evaluated and the RELAP5-3D has been successfully modified to fulfill the associated requirements. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effect experiments were adopted. Through the assessments against separate-effect experiments, the success of the code modification in accordance with Appendix K of 10 CFR 50 was demonstrated. Another six sets of integral-effect experiments will be applied in the next step to assure the integral conservatism of the Appendix K version of RELAP5-3D on LOCA licensing evaluation.     

燃料热导率降级对CAP1000大破口失水事故的影响分析

在高燃耗情况下,燃料芯块的热导率随燃耗降低,该现象被称之为热导率降级(TCD)现象。TCD现象影响失水事故(LOCA)前稳态工况的燃料平均温度和燃料储能,进而影响大破口LOCA过程中的包壳峰值温度(PCT)。本研究采用大破口LOCA分析程序WCOBRA/TRAC对CAP1000冷段双端剪切断裂事故进行了不同燃耗的敏感性分析,并获得了不同工况下的PCT。分析中采用美国核燃料研究所(NFI)修正的TCD模型对降级后的燃料热导率进行模拟,同时考虑了燃耗大于30GW·d/tU后FQ和FΔh峰值因子的降低。敏感性分析表明,考虑TCD和峰值因子降低的影响,PCT极限工况不再出现在低燃耗区间,而出现在燃耗为29GW·d/tU附近。与其他燃耗水平相比,该燃耗点的PCT第1峰值和第2峰值均处于最高水平。本研究结果可为高燃耗情况下非能动电厂大破口LOCA的分析评估提供参考。     

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 analysis of the power curve sensitivity test series at ROSA-III

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR/6) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. Seven recirculation pump suction line break LOCA experiments were conducted at the ROSA-III facility in order to examine the effect of the initial stored heat of a fuel rod on the peak cladding temperature (PCT). The break size was changed from 200% to 5% in the test series and a failure of a high pressure core spray (HPCS) diesel generator was assumed. Three power curves which represented conservative, realistic and zero initial stored heat, respectively, were used.In a large break LOCA such as 200% or 50% breaks, the initial stored heat in a fuel rod has a large effect on the cladding surface temperature because core uncovery occurs before all the initial stored heat is released, whereas in a small break LOCA such as a 5% break little effect is observed because core uncovery occurs after the initial stored heat is released. The maximum PCTs for the conservative initial stored heat case was 925 K, obtained in the 50% break experiment, and that for the realistic initial stored heat case was 835 K, obtained in the 5% break experiment.     

Application of the COBRA/TRAC best-estimate methodology to the AP600 large-break LOCA analysis

The AP600 is a simplified advanced pressurized water reactor (PWR) design incorporating passive safety systems that perform the same function as the active emergency core cooling systems (ECCSs) on the current reactors. In order to verify the effectiveness of the AP600 design features for mitigation of a postulated large-break loss-of-coolant accident (LOCA), the recently United States Nuclear Regulatory Commission (USNRC)-approved best-estimate LOCA methodology (BELOCA) was applied to perform the AP600 standard safety analysis report large-break LOCA analysis. The applicability of the COBRA/TRAC code to model the AP600 unique features was validated against cylindrical core test facility (CCTF) and upper plenum test facility (UPTF) downcomer injection tests, the blowdown and reflood cooling heat transfer uncertainties were re-assessed for the AP600 large-break LOCA conditions and a conservative minimum film boiling temperature was applied as a bounded parameter for blowdown cooling. The BELOCA methodology was simplified to quantify the code uncertainties due to local and global models, as well as the statistical approximation methods, with the other uncertainties being bounded by limiting assumptions on the initial and boundary conditions. The final 95th percentile peak cladding temperature (PCT^95%) was 1186 K, which meets the 10CFR50.46 criteria with a considerable margin. It is therefore concluded that the AP600 design is effective in mitigation of a postulated large-break LOCA.