Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

Pressurized heavy water reactor fuel behaviour in power ramp conditions

In order to check and improve the quality of the Romanian CANDU fuel, an assembly of six CANDU fuel rods has been subjected to a power ramping test in the 14 MW TRIGA reactor at INR. After testing, the fuel rods have been examined in the hot cells using post-irradiation examination (PIE) techniques such as: visual inspection and photography, eddy current testing, profilometry, gamma scanning, fission gas release and analysis, metallography, ceramography, burn-up determination by mass spectrometry, mechanical testing. This paper describes the PIE results from one out of the six fuel rods. The PIE results concerning the integrity, dimensional changes, oxidation, hydriding and mechanical properties of the sheath, the fission-products activity distribution in the fuel column, the pressure, volume and composition of the fission gas, the burn-up, the isotopic composition and structural changes of the fuel enabled the characterization of the behaviour of the Romanian CANDU fuel in power ramping conditions performed in the TRIGA materials testing reactor.     

A qualitative identification and analysis of hazards,risks and operating procedures for a decommissioning safety assessment of a nuclear research reactor

Korea Research Reactor-1(KRR-1, TRIGA Mark-II type reactor), the first nuclear research reactor in Korea, is being prepared for a decommissioning. The decommissioning methods and procedures of KRR-1 ought to be based on its structural conditions and radiological characteristics. Also, a systematic approach to the decommissioning tasks must be followed by reviews and assessments of the decommissioning workers’ safety.     

基于PSA的非能动余热排出系统可靠性研究

非能动型反应堆概率安全评价（PSA）工作在分析非能动系统可靠性时,仅考虑系统设备可靠性,未涉及物理过程可靠性。综合考虑非能动系统设备可靠性与物理过程可靠性时,又存在仅考虑系统投入的设备可靠性而忽略运行设备可靠性的问题。针对此问题,以丧失正常给水事故下AP1000非能动余热排出系统（PRHRS）为研究对象,采用自主提出的综合法将系统可靠性融合进PSA模型,兼顾能动设备的需求失效与非能动设备的运行失效,分析了系统设备可靠性的敏感性。结果表明,综合法对PRHRS进行可靠性分析时所得事故序列谱更真实、更全面,与传统方法相比较具有优越性。      

岭澳核电站3,4号机组反应堆保护系统概率安全评价结果校算

为了评估数字化仪表控制系统对核电厂安全的影响,以电厂停堆系统和专设安全设施驱动系统为例,参考西门子公司提供的故障树逻辑,对主泵流量低及功率量程中子通量高于整定值停堆故障和蒸汽发生器(SG)低-低水位和同一SG中主给水流量低故障进行了概率安全分析.分析中分别采用西门子公司提供的输入数据及通过失效率、试验时间以及β因子方法计算得到的数据,对西门子的分析结果进行了校算,在主要割集和失效概率上得到更为真实的结果.结果表明,考虑2种多样性的反应堆保护系统停堆I&C功能需求失效概率均值为5.5×10~(-8),符合分布式控制系统(DCS)合同中确定的可靠性目标值(1.0×10~(-7))和辅助给水电动泵驱动信号功能需求失效概率均值(5.21×10~(-6)与8.32×10~(-6)),也符合DCS合同中确定的可靠性目标值(1.0×10~(-5)).     

功能可靠性分析方法在XAPR堆芯自然循环可靠性评价中的应用

功能失效是导致自然循环系统运行失效的重要因素,需要在其可靠性分析中予以考虑。基于功能可靠性评价流程,通过RELAP5程序模拟自然循环物理过程,对西安脉冲堆（XAPR）池水自然循环冷却堆芯能力的可靠性进行评价。结合中破口失水事故,根据包壳完整性的功能准则,确定影响自然循环的关键参数;采用拉丁超立方抽样确定输入参数组合,进行参数敏感性分析和功能可靠性评价,并将功能可靠性评价结果整合到概率安全评价（PSA）模型中。分析结果表明:在PSA模型中不仅需要考虑硬件可靠性,还应充分考虑功能可靠性,以更好地指导XAPR运行及提高其安全性。      

Reliability assessment of passive containment isolation system using APSRA methodology

In this paper, a methodology known as APSRA (Assessment of Passive System ReliAbility) has been employed for evaluation of the reliability of passive systems. The methodology has been applied to the passive containment isolation system (PCIS) of the Indian advanced heavy water reactor (AHWR). In the APSRA methodology, the passive system reliability evaluation is based on the failure probability of the system to carryout the desired function. The methodology first determines the operational characteristics of the system and the failure conditions by assigning a predetermined failure criterion. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the PCIS performance. APSRA proposes to compare the code predictions with the test data to generate the uncertainties on the failure parameter prediction, which is later considered in the code for accurate prediction of failure surface of the system. Once the failure surface of the system is predicted, the cause of failure is examined through root diagnosis, which occurs mainly due to failure of mechanical components. The failure probability of these components is evaluated through a classical PSA treatment using the generic data. The reliability of the PCIS is evaluated from the probability of availability of the components for the success of the passive containment isolation system.     

Qualification Tests for Packages used for Transport and Storage of Radioactive Waste (Low Activity) in INR Pitesti

Abstract

Radioactive wastes generated by the TRIGA INR research reactor are packaged according to the national and international standards and the IAEA Regulations. The technology for packaging and treatment of radioactive wastes used in this institute can be applied, prospectively, at the Nuclear Power Plant Cernavoda, after commissioning. The qualification tests (type tests) are described for packages used for transport and storage (for a long period of about 30 years) of radioactive wastes (low activity, up to 0.5068×10¹⁰ Bq per drum, or 0.164 Ci per drum). The package used is a drum manufactured in Romania industry. The type tests carried out are described i.e. compression, penetration, free fall, leaching, safety in use (biological protection), chemical and mechanical characteristics, effect on the environment, and also the results, interpretation and conclusions from the tests. As a result of the tests, Romanian technology for treatment and packaging of radioactive wastes is considered to be in accordance with IAEA Regulations.     

Passive system reliability in the nuclear power plants (NPPs) using statistical modeling

The probabilistic safety assessment (PSA) has been studied for the very high temperature reactor (VHTR). There is a difficulty to make the quantification of the PSA due to the deficiency of the operation and experience data. So, it is necessary to use the statistical data for the basic event. The physical data of the non-linear fuzzy set algorithm are used to quantify the designed case. The mass flow rate in natural circulation is investigated. In addition, the potential energy in the gravity, the temperature and pressure in the heat conduction, and the heat transfer rate in the internal stored energy are investigated. The values in the probability set and fuzzy set are compared for the failure explanation. The result shows how to use the newly made probability of the failure in the propagations. The failure frequencies, which are made by the GAMMA (GAs Multi-component Mixture Analysis) code, are compared with four failure frequencies by probabilistic and fuzzy methods. The results show that the artificial intelligence analysis of the fuzzy set could improve the reliability method than that of the probabilistic analysis.     

Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes

The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR–R1 reactor are calculated. The idea is to obtain the systematic error for k_∞ for this methodology comparing the calculated and the experimental results.     

Fuzzy probability on reliability study of nuclear power plant probabilistic safety assessment: A review

Fault tree analysis (FTA) is a graphical model which has been widely used as a deductive tool for nuclear power plant (NPP) probabilistic safety assessment (PSA). The conventional one assumes that basic events of fault trees always have precise failure probabilities or failure rates. However, in real-world applications, this assumption is still arguable. For example, there is a case where an extremely hazardous accident has never happened or occurs infrequently. Therefore, reasonable historical failure data are unavailable or insufficient to be used for statistically estimating the reliability characteristics of their components. To deal with this problem, fuzzy probability approaches have been proposed and implemented. However, those existing approaches still have limitations, such as lack of fuzzy gate representations and incapability to generate probabilities greater than 1.0E-3. Therefore, a review on the current implementations of fuzzy probabilities in the NPP PSA is necessary. This study has categorized two types of fuzzy probability approaches, i.e. fuzzy based FTA and fuzzy hybrid FTA. This study also confirms that the fuzzy based FTA should be used when the uncertainties are the main focus of the FTA. Meanwhile, the fuzzy hybrid FTA should be used when the reliability of basic events of fault trees can only be expressed by qualitative linguistic terms rather than numerical values.     

Experimental validation of the thermal-hydraulic code SACATRI

A sub-channel analysis steady state thermal-hydraulic code (SACATRI) was developed for the Moroccan TRIGA MARK II research reactor. The main objective of the thermal-hydraulic study of the whole reactor core is to evaluate the main safety parameters of the reactor core, and to ensure that they are within the safety limits for any operating conditions. The thermal-hydraulic model used in SACATRI is based on four partial differential equations that describe the conservation of mass, energy and momentum. In order to assess the thermal-hydraulic mathematical model of SACATRI, the present paper focuses on the quantification of the physical model accuracy to judge if the code is capable to represent the thermal-hydraulic behaviour of the reactor core with sufficient accuracy. The methodology adopted is based on the comparison between responses from SACATRI computational model and experimentally measured responses performed on the IPR-R1 TRIGA research reactor. The results showed good agreement between SACATRI predictions and the experimental measurements where the discrepancies observed (simulation-experiment) are less than 6%.     

多模块高温气冷堆地震概率安全分析关键技术研究

高温气冷堆核电厂采取多个反应堆模块匹配1个汽轮机的设计方式,即1台高温气冷堆机组会包含多个反应堆模块,这使多个高温气冷堆模块在地震外部事件下存在明显的相关性,因此在利用概率风险分析方法来全面地识别和评价高温气冷堆的地震风险时,需要从机组的角度充分考虑和模化机组内多个反应堆模块间的相关性。高温气冷堆示范电站已完成了较为完整的单模块地震概率安全分析,本文将以该分析结果为基础梳理出高温气冷堆多模块地震概率安全分析的关键技术要素并进行研究,研究内容包括多模块事件序列建模和地震相关性失效评价等关键技术,并针对多模块高温气冷堆提出了应用策略。然后以双模块设计的高温气冷堆示范电站为对象,以地震导致丧失厂外电始发事件为代表,对多模块高温气冷堆地震概率安全分析进行了实例分析获得远低于概率安全目标的释放类频率,且分析得到了高温气冷堆多模块事件序列建模策略与地震相关性失效的评价路线可行这一重要结论。     

HFETR供电系统概率安全评价

反应堆供电系统失效可导致堆芯熔毁等严重事故后果。本工作应用RiskSpectrum软件,对高通量工程试验堆(简称HFETR)供电系统开展概率安全评价(PSA)工作。通过整合部分法考虑共因故障,建立了以全场断电(SBO)为顶事件的系统故障树模型,并定量给出HFETR发生SBO概率为7.49×10~(-8),证明HFETR现役供电系统安全可靠。同时,以供电系统模型及运行可靠性数据为基础,进行了割集、重要度、敏感度等分析,较全面地分析了现役供电系统的风险水平,为HFETR供电系统变更、升级和改造提供了重要参考。     

A study of the decommissioning procedure of an activated structure through an evaluation of the decommissioning cost for a research reactor

As decommissioning of a research reactor and a nuclear installation requires a long period of time from the decommissioning preparation work to the site remediation, the management of the data generated during the entire period of decommissioning is one of the most important tasks. In particular, the data obtained from research reactor decontamination and decommissioning activities can be important resources securing the safety and economic feasibility for other research reactor decommissioning. The owner of the research reactor and nuclear power plant need to submit decommissioning plan to the regulatory body at the starting stage of the research reactor and nuclear installation decommissioning project. The cost plan for decommissioning and the method for assessing the amount of exposure to protect workers must be stated in the decommissioning plan.This paper introduces the DES (Decommissioning Engineering System) that can be able to manage the data generated in the process of decommissioning of the TRIGA research reactor, to calculate an amount of waste, to evaluate decommissioning cost after deriving unit work productivity factor, and to predict the decommissioning process in advance. To verify the usability of this system and data integrity through connections among the unit systems, it describes the process to calculate the decommissioning cost using the data generated in dismantling an activated bio-shielding concrete in the TRIGA research reactor.As a result of the experiment to calculate the decommissioning cost with the TRIGA research reactor structure, it was found that the calculations were done precisely without flaw as the purpose of the experiment. Therefore, the DES can not only be used for other research reactors decommissioning, but also it is expected to be applied to other research reactors in the future. As a decommissioning cost between an activated concrete and a non-activated concrete according to the method of the dismantling procedure was significantly different, a study regarding the dismantling procedure needs more research.     

Framework analysis of fluoride salt-cooled high temperature reactor probabilistic safety assessment

Probabilistic safety assessment(PSA) is important in nuclear safety review and analysis. Because the design and physics of the fluoride salt-cooled high temperature reactor(FHR) differ greatly from the pressurized water reactor(PWR), the methods and steps of PSA in FHR should be studied. The high-temperature gascooled reactor(HTR-PM) and sodium-cooled fast reactors have built the PSA framework, and the framework to finish the PSA analysis. The FHR is compared with the PWR, HTR-PM and sodium-cooled fast reactors from the physics, design and safety. The PSA framework of FHR is discussed. In the FHR, the fuel and coolant combination provides large thermal margins to fuel damage(hundreds of degrees centigrade). The tristructuralisotropic(TRISO) as the fuel is independent in FHR core and its failure is limited for the core. The core damage in Level 1 PSA is of lower frequency. Levels 1 and 2 PSA are combined in the FHR PSA analysis. The initiating events analysis is the beginning, and the source term analysis and the release types are the target. Finally, Level3 PSA is done.     

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C,JENDL-3.3 and ENDF/B-Vl data libraries

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for ^natZr, ^natMo, ^natCr, ^natFe, ^natNi, ^natSi, and ^natMg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ²⁸, δ²⁵, ρ²⁵, and C¹ were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries.     

西安脉冲堆满功率运行工况内部始发事件一级概率安全评价

针对西安脉冲堆(XAPR)2 MW满功率运行工况,建立了内部始发事件一级概率安全评价(PSA)模型,对始发事件识别、事故序列分析及可靠性数据处理等进行了研究。应用小事件树-大故障树方法,在Risk Spectrum平台上完成XAPR堆芯损伤事故序列的定量分析。结果表明,XAPR内部事件导致的堆芯损伤频率(CDF)为4.14×10~(-6)/(堆·年),对CDF贡献最大的为堆水池堆芯高度处大破口失水事故,支配性事故序列是大破口失水事故后紧急排水系统失效。研究结果证明XAPR具有较高的安全性。     

反应堆压力容器承压热冲击中的PSA方法研究

在瞬态过程中,当处于承压状态下的反应堆压力容器(RPV)的内表面被快速冷却时,即为承压热冲击(PTS)。由此,反应堆压力容器可能出现贯穿裂纹而失效。为分析PTS事件导致RPV出现裂纹的频率,需要进行概率安全评价(PSA)。通过PSA模型确定可能引起PTS的事件序列,并结合这些序列的热工水力分析结果,为PTS概率断裂力学分析提供支持。     

贝叶斯估计方法在Alpha因子模型参数估计中的应用

核电厂的概率安全分析（PSA）结果表明,共因失效（CCF）在系统的不可靠度中占有相当重要的贡献。国内PSA分析中CCF数据一直采用通用数据,难以体现国内核电机组的运行特点。Alpha因子模型由于其参数估计的简单化、计算结果的精确性等特点是PSA中最常用于模化共因失效的模型。但由于共因失效事件的罕见性,使用经典估计算法难以产生合理的统计值,因此,本文给出共因参数的贝叶斯估计方法,该方法能够结合先验信息和样本信息,不需要很大的样本就能得到较好的估计值,有效解决了核电厂共因失效事件少、使用经典估计方法计算结果不合理的问题,适用于核电厂共因失效模型参数估计。