实验室通风空调设计浅谈

核电厂作为一个大型的工业体系，其实验室的功能与民用类建筑迥异，为了保障实验人员的安全与健康。使科研人员有良好的操作条件，以促进研究工作的顺利开展，必须合理地设计实验室通风空调系统。本文根据核电厂中实验室通风空调设计的特点，详细论述了设计目标和方法，从而为未来核电厂中实验室通风空调设计提供了参考。     

碘吸附器在核电厂通风系统中的应用

碘吸附器是核电厂通风系统中普遍使用的用于保障在事故工况下主控室可居留环境和减少排风放射性水平以满足排放限值要求的部件。它对核电厂事故产物中放射性碘元素的吸附效率或净化系数直接影响到相关安全分析的结论;碘吸附器在通风系统中的防火设计也是核电厂防火系统安全审评中关注的重要问题。本文阐述了碘吸附器在核电厂应用中需要考虑的问题...     

EPR系统安全一致性评价方法研究

系统安全一致性评价是一种确定论安全评价方法,对核电厂系统设计是否满足国家相关法规以及设计总体阶段所制定的安全要求进行评价,是确保核电厂安全的重要保障。系统安全一致性评价是找出系统设计薄弱环节的重要途径,进而指导系统改进设计。本文根据工程实践归纳出系统安全一致性评价方法与思路,并结合实例对评价过程进行说明。该方法已运用在三代核电EPR设计中,并已完成主要核岛系统安全一致性评价工作。该方法可供CPR1000核电厂设计改进参考,也适合运用于定期安全审查(PSR)工作。     

机械通风冷却塔在内陆核电厂中的应用

介绍了如何将核电厂相关设计规范应用于内陆核电厂最终热阱输热系统的机械通风冷却塔的等级划分,分析和总结了核级机械通风冷却塔的特殊设计和建造要求,并提出了核级机械通风冷却塔的设备鉴定过程,为我国自行设计和开发核级机械通风冷却塔提供借鉴和指导作用.     

CANDU6核电厂无过滤安全壳通风模式的研究

CANDU6核电厂早期设计未考虑严重事故对策,在严重事故下,CANDU6核电厂的安全壳容易失效。为了解决这一问题,本文研究了无过滤安全壳通风模式对CANDU6核电厂安全壳的影响。本文选取典型的全厂断电严重事故,利用重水蒸气回收系统作为无过滤安全壳通风的路径,初步研究了该通风模式下对安全壳完整性的保持和对裂变产物源项的滞留能力。研究表明:该通风模式可以有效保持安全壳的完整性,同时,对裂变产物源项也有一定的滞留能力。     

核电厂电气二次系统信息安全监管平台的研究

本文介绍了核电厂电气二次系统技术的发展趋势及其面临的网络安全威胁问题;分析了国家能源局对核电厂电气二次系统安全防护部署的强制性技术监管要求;通过解析核电厂电气二次系统的设备现状,研究出一种可实际部署在核电厂的电气二次系统的信息安全监管平台方案,并进一步探讨了该方案的后续发展趋势。     

《核电厂可行性研究阶段厂址安全分析技术规范》等2项标准通过审查

正能源行业核电标准化技术委员会秘书处于2014年11月19日~20日在北京组织召开核电标准审查会,本次会议审查了由中国核电工程有限公司主编的《核电厂可行性研究阶段厂址安全分析技术规范》和河南核净洁净技术有限公司主编的《核空气和气体处理规范通风、空调与空气净化第16部分:净化部件用排架》。来自环境保护部核与辐射安全中心、中核核电运行管理有限公司、山东     

秦山第二核电厂通风过滤器运行管理

简要介绍了核电厂几种常用高效过滤器的结构,分析了过滤器压差测量不准确的原因以及采取的纠正措施,总结了秦山第二核电厂通风过滤器投入运营以来遇到的一些问题,对典型实例进行了剖析和经验反馈。     

核电厂数据通信系统的安全审评

介绍了核电厂数据通信系统安全审评的目的和主要依据,简要论述了核安全审评中需要关注的重点问题,分析了可能影响到它所支持的系统执行所要求的安全功能的一些因素。     

大温差冷冻水系统在核电厂中的应用研究

文章从运行能耗、核电厂核岛厂房布置及初投资三方面分析核岛厂房内空调冷冻水系统采用大温差设计的意义。在理论分析的基础上,通过公式推导计算,探讨大温差冷冻水系统相对于常规冷冻水系统在运行节能方面的优势。结合低温送风技术分析大温差技术对降低核电厂空调系统的初投资和改善核岛厂房布置的重要意义。     

HTR-10安全负压通风系统的设计

由于高温气冷堆具有固有安全性，因此在高温气冷堆的设计中采用通风式包容体替代了密封承压式安全壳，在供暖，通风与空调（HVAC）系统中相应采用了安全负压通风系统，以保证包容体在正常工况或事故工况下都能满足与安全相关的一切功能，本文介绍了10MW高温气冷实验堆（HTR－10）安全负压通风系统的设计与评价。     

Assessment of control room habitability and unfiltered air inleakage test of the OPR 1000 NPP

The assessment of control room habitability for Hanbit unit 5 was performed based on GL 2003-01 and Regulatory Guide 1.197. The integrated control room envelope (CRE) test was performed utilizing ASTM E741. Four tests were performed using each of the control room HVAC subtrains. The control room heating, ventilating, and air conditioning (HVAC) system lineup of pressurization mode test was based upon a lineup that encompassed the design basis radiological analyses. The other control room HVAC system lineup of isolation mode test was based on an operation mode that considers toxic gas. The measured inleakage for the isolation test mode remains within the toxicity limit. Radiation effect analysis showed that the measured inleakage satisfied the regulatory criteria, and the inleakage would not result in control room operator dose exceeding 50 mSv whole body and 500 mSv thyroid except train A pressurization test mode. The thyroid dose due to maximum measured unfiltered inleakage of 8976 lpm for train A is corresponding to 700 mSv. Modifications to the CRE boundary and control room HVAC system should be done to demonstrate that the measured unfiltered inleakage for train A pressurization test mode is bounded by the regulatory criteria assumed in the design basis radiological analyses.     

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR.

For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems.     

Issues affecting advanced passive light-water reactor safety analysis

Next generation commercial reactor designs emphasize enhanced safety by means of improved safety system reliability and performance. These objectives are achieved via safety system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet, the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs will necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing U.S. advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes may require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.     

Development of the FHR advanced natural circulation analysis code and application to FHR safety analysis

The University of California, Berkeley (UCB) is performing thermal hydraulics safety analysis to develop the technical basis for design and licensing of fluoride-salt-cooled, high-temperature reactors (FHRs). FHR designs investigated by UCB use natural circulation for emergency, passive decay heat removal when normal decay heat removal systems fail. The FHR advanced natural circulation analysis (FANCY) code has been developed for assessment of passive decay heat removal capability and safety analysis of these innovative system designs. The FANCY code uses a one-dimensional, semi-implicit scheme to solve for pressure-linked mass, momentum and energy conservation equations. Graph theory is used to automatically generate a staggered mesh for complicated pipe network systems. Heat structure models have been implemented for three types of boundary conditions (Dirichlet, Neumann and Robin boundary conditions). Heat structures can be composed of several layers of different materials, and are used for simulation of heat structure temperature distribution and heat transfer rate. Control models are used to simulate sequences of events or trips of safety systems. A proportional-integral controller is also used to automatically make thermal hydraulic systems reach desired steady state conditions. A point kinetics model is used to model reactor kinetics behavior with temperature reactivity feedback. The underlying large sparse linear systems in these models are efficiently solved by using direct and iterative solvers provided by the SuperLU code on high performance machines. Input interfaces are designed to increase the flexibility of simulation for complicated thermal hydraulic systems. This paper mainly focuses on the methodology used to develop the FANCY code, and safety analysis of the Mark 1 pebble-bed FHR under development at UCB is performed.     

Power-generating unit with dual-loop sodium-cooled fast reactor and dual-wall steam generator

The results of a probabilistic analysis performed to validate the safety of AES-2006 designed for the site of the Novovoronezh nuclear power plant are presented. The requirements for the AES-2006 design are examined. The characteristic features of the AES-2006 design for the conditions at the Novovoronezh nuclear power plant site are described, including the diversity of the equipment and operating regime, passive systems, and scheduled maintenance of safety systems with the reactor operating at power. The scope of the probabilistic safety analysis performed at the development stage of the technical design is described. The important problems which must be solved in a probabilistic safety analysis for the designs of new nuclear power plants are discussed. Translated from Atomnaya énergiya,Vol. 106, No. 3, pp. 123–129, March, 2009.     

Modular HTR confinement/containment and the protection against aircraft crash

Does an HTR need a containment – pressure resistant – or is it possible – licensable – to have only a so-called confinement.The answer depends on both the results of the safety analysis of the accidents considered in the design and the acceptance by the licensing authorities and the public of a safety approach only based on severe core damage exclusion.The safety approach to be developed for modular HTRs must describe the application of the defence in depth principle for such reactors. Whatever the requirements on the last confinement barrier could be, a convincing demonstration of the exclusion of any severe core damage is needed, relying on exhaustive and bounding considerations of severe core damage initiators and the use of non-questionable arguments.The paper presents the containment issues for HTRs based on German experience background and considerations for modern modular HTR safety approach including beyond design situations.

• For the German HTRs (designed in the 80s), it could be shown in the licensing procedures in Germany that there was no need for a pressure retaining and gas tight containment to enclose radioactive nuclides released from the nuclear heat source. Instead, the confinement envelope acted in conjunction with other barriers to minimize the release of radioactive nuclides and the radiological impact on the environment.

• The confinement envelope consisted of the reactor building, a sub-atmospheric pressure system, a building pressure relief system, an HVAC systems isolation and a filtration system.

• During a major depressurization accident, unfiltered releases were discharged to the environment. The analyses results show that the environmental impact was far below the dose limits according to the German Radiological Protection Ordinance, even when the effect of filters was not taken into account.

• The demonstration strongly relied on the assumptions made for the source term definition, e.g. the fuel particles failure rates (under irradiation and during accidental conditions), the diffusion data, the dust data and the deposition/lift-off mechanisms.

• For modern modular HTRs, the last confinement barrier performances will have to be determined in accordance with the set of accidents to be considered in the design including internal and external hazards and the limits targeted for the public and the environment protection.

Further more the paper presents an analysis of effects of a deliberate crash of a large commercial airliner on a former German HTR design.     

Functional reliability analysis of Safety Grade Decay Heat Removal System of Indian 500 MWe PFBR

Passive systems are increasingly deployed in nuclear industry with an objective of increasing reliability and safety of operations with reduced cost. Methods for assessing the reliability of thermal-hydraulic passive systems, that is systems with moving working fluid, address the issues in natural buoyancy-driven flow that could result in a failure to meet the design safety limits under accident scenarios. This is referred as design functional reliability. This paper presents the results of functional reliability analysis carried out for the passive Safety Grade Decay Heat Removal System (SGDHRS) of Indian Prototype Fast Breeder Reactor (PFBR). The analysis is carried out based on the overall approach reported in the Reliability Methods for Passive System (RMPS, European Commission) project. Functional failure probability is calculated using Monte-Carlo method and also with method of moments.     

Fuzzy uncertainty modeling applied to AP1000 nuclear power plant LOCA

This article presents an uncertainty modeling study using a fuzzy approach applied to the Westinghouse advanced nuclear reactor. The AP1000 Westinghouse Nuclear Power Plant (NPP) is provided of passive safety systems, based on thermo physics phenomenon, that require no operating actions, soon after an incident has been detected. The use of advanced passive safety systems in NPP has limited operational experience. As it occurs in any reliability study, statistically non-significant events report introduces a significant uncertainty level about the failure rates and basic events probabilities used on the fault tree analysis (FTA). In order to model this uncertainty, a fuzzy approach was employed to reliability analysis of the AP1000 large break Loss of Coolant Accident (LOCA). The final results have revealed that the proposed approach may be successfully applied to modeling of uncertainties in safety studies.