Experimental study on the sub-atmospheric loop heat pipe passive cooling system for spent fuel pool

As one kind of the natural circulation cooling system, loop heat pipe is promising in improving the safety of the nuclear power station since it is passive and has no electricity driven components. A novel heat pipe cooling system is designed for passively removing the residual heat released by the spent fuel stored in the spent fuel pool (SFP) under the accidental conditions such as the station blackout. This system is characterized by its large-diameter and long-length evaporator. Its working fluid is water and it's sub-atmospheric. To test such system's heat transfer performance and get to know its thermo-fluid dynamics, a test facility for a simplified heat pipe made of one evaporator tube and one condenser has been developed. The heat transfer rate of the simplified heat pipe is obtained in a wide range of conditions covering the potential working conditions in spent fuel pool. Moreover, it's found that heat pipe with such a large-diameter and long-length evaporator is vulnerable to be unstable. The periodic state mode is more likely to happen when the heat source temperature, the air velocity or the volumetric filling ratio is low. Furthermore, the effects of hot water temperature, the air velocity and the filling ratio of the water in the circulation system have been analyzed.     

National Ignition Facility TestController for automated and manual testing

The Controls and Information Systems (CIS) organization for the National Ignition Facility (NIF) has developed controls, configuration and analysis software applications that combine for several million lines of code. The team delivers updates throughout the year, from major releases containing hundreds of changes to patch releases containing a small number of focused updates. To ensure the quality of each delivery, manual and automated tests are performed using the NIF TestController test infrastructure. The TestController system provides test inventory management, test planning, automated and manual test execution, release testing summaries and results search, all through a web browser interface. As part of the three-stage software testing strategy, the NIF TestController system helps plan, evaluate and track the readiness of each release to the NIF production environment.After several years of use in testing NIF software applications, the TestController's manual testing features have been leveraged for verifying the installation and operation of NIF Target Diagnostic hardware. The TestController recorded its first test results in 2004. Today, the system has recorded the execution of more than 160,000 tests and continues to play a central role in ensuring that NIF hardware and software meet the requirements of a high reliability facility. This paper describes the TestController system and discusses its use in assuring the quality of software delivered to the NIF.     

Internet accessible hot cell with gamma spectroscopy at the Missouri S&T nuclear reactor

A dual-chambered internet-accessible heavily shielded facility with pneumatic access to the University of Missouri Science and Technology (Missouri S&T) 200 kW Research Nuclear Reactor (MSTR) core has been built and is currently available for irradiation and analysis of samples. The facility allows authorized distance users engaged in collaborative activities with Missouri S&T to remotely manipulate and analyze neutron irradiated samples. The system consists of two shielded compartments, one for multiple sample storage, and the other dedicated exclusively for radiation measurements and spectroscopy. The second chamber has multiple detector ports, with graded shielding, and has the capability to support gamma spectroscopy using radiation detectors such as an HPGe detector. Both these chambers are connected though a rapid pneumatic system with access to the MSTR nuclear reactor core. This new internet-based system complements the MSTR's current bare pneumatic tube (BPT) and cadmium lined pneumatic tube (CPT) facilities. The total transportation time between the core and the hot cell, for samples weighing 10 g, irradiated in the MSTR core, is roughly 3.0 s. This work was funded by the DOE grant number DE-FG07-07ID14852 and expands the capabilities of teaching and research at the MSTR. It allows individuals who do not have on-site access to a nuclear reactor facility to remotely participate in research and educational activities.     

The Test Facility for the EAST Superconducting Magnets

A large facility for testing superconducting magnets has been in operation at the Institute of Plasma Physics of the Chinese Academy of Sciences since the completion of its construction that began in 1999. A helium refrigerator is used to cool the magnets and liquefy helium which can provide 3.8 K-4.5 K, 1.8 bar-5 bar, 20g/s-40g/s supercritical helium for the coils or a 150 L/h liquefying helium capacity. Other major parts include a large vacuum vessel （3.5 m in diameter and 6.1 m in height） with a liquid nitrogen temperature shield, two pairs of current lead, three sets of 14.5 kA-50 kA power supply with a fast dump quench protection circuitry, a data acquisition and control system, a vacuum pumping system, and a gas tightness inspecting devise. The primary goal of the test facility is to test the EAST TF and PF magnets in relation to their electromagnetic, stability, thermal, hydraulic, and mechanical performance. The construction of this facility was completed in 2002, followed by a series of systematic coil testing. By now ten TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a model coil of the PF large coil have been successfully tested in the facility.     

Preliminary design of the control and data acquisition systems for the Neutral Beam Test Facility

The Neutral Beam Test Facility, which will be built in Padova, Italy, is aimed at developing the ITER heating neutral beam injector (HNB) and at testing and optimizing its operation up to nominal performance before installation on ITER. It requires the development of two independent experiments referred to as SPIDER (source for production of ions of deuterium extracted from Rf plasma) and MITICA (megavolt ITer injector & concept advancement). SPIDER will explore the full-size negative ion source for ITER, whereas MITICA will explore the full-size ITER neutral beam injector. Both experiments will be designed for long-pulse operation, up to 3600 s, as ITER itself. MITICA includes three functional components: the heating neutral beam injector plant system (HNB), which is the device under test; the auxiliary plant system (AUX), which includes all equipment to operate the HNB in the test facility (e.g. the local electric grid to feed the HNB power supplies), and MITICA supervisory system that is an electronics/informatics infrastructure to operate the facility. The paper introduces the requirements for the control and data acquisition systems of the experiments and proposes a preliminary design for both systems. SPIDER, which is preparatory to MITICA and will be developed on a shorter time scale, has no constraints coming from ITER CODAC, whereas MITICA includes the ITER neutral beam injector and therefore must be fully compatible with ITER CODAC.     

AC－600PWR蒸汽发生器模拟体设计

介绍了ＡＣ－６００ＰＷＲ二次侧非能动应急堆芯余热排出系统整体效应实验装置中蒸汽发生器（ＳＧ）模拟体的模拟准则和主要设计特点。其主要模拟准则为功率－容积比准则，但在尽量保证再现原型主要热工水力过程的前提下，进行了适当的修改和简化．该模拟体总体比例为１／３９０，换热管束及内件可更换，以满足不同实验研究要求。     

ACME整体性能试验设施工作压力选取方案分析

拟建造的先进堆芯冷却机理试验台架（ACME）是验证压水堆核电站非能动安全系统性能及其安全分析软件的整体性能试验设施。在介绍AP1000电站整体性能试验台架及其评价的基础上，分析了不同工作压力对试验的影响。重点阐述了ACME工作压力的选取方案及其特点，探讨了确定试验初始状态的方法。分析表明：选取9.3MPa作为ACME的工作压力，涵盖了主要非能动系统工作的压力范围，具有等压等物性模拟非能动压水堆电站LOCA等事故工况的能力，是一个先进的非能动堆芯冷却整体性能试验设施设计方案。     

ITER articulated inspection arm (AIA): R&d progress on vacuum and temperature technology for remote handling

This paper is part of the remote handling (RH) activities for the future fusion reactor ITER. The aim of the R&D program performed under the European Fusion Development Agreement (EFDA) work program is to demonstrate the feasibility of close inspection tasks such as viewing or leak testing of the Divertor cassettes and the Vacuum Vessel (VV) first wall of ITER.It is assumed that a long reach, limited payload carrier penetrates the ITER chamber through the openings evenly distributed around the machine such as In-Vessel Viewing System (IVVS) access or through upper port plugs.To perform an intervention a short time after plasma shut down, the operation of the robot should be realised under ITER conditioning i.e. under high vacuum and temperature conditions (120 °C).The feasibility analysis drove the design of the so-called articulated inspection arm (AIA) which is a 8.2 m long robot made of five modules with a 11 actuated joints kinematics. A single module prototype was designed in detail and manufactured to be tested under ITER realistic conditions at CEA-Cadarache test facility.As well as demonstrating the potential for the application of an AIA type device in ITER, this program is also dedicated to explore the necessary robotic technologies required to ITER's IVVS deployment system.This paper presents the whole AIA robot concept, the first results of the test campaign on the prototype vacuum and temperature demonstrator module.     

Design and Testing of the Nupak-200 Type-B(U)F Packaging

Abstract

The design and testing of the Nupak-200 Type-B(U)F packaging developed by AEA Technology with support from the United Kingdom Atomic Energy Authority (UKAEA) is described. The design concept and operational criteria are discussed, and the packaging's principal features are reviewed. The adaptability of the two-box assembly is examined, as is its application to current and future radioactive materials transport requirements. The paper highlights the test programme that was devised to simulate the impalement of composite steel and cork panels on a punch of the type used for testing to IAEA standards. This was done in order to generate adequate data about the materials' characteristics for the finite analysis programs which supported the design. The paper concludes by reviewing the drop-test. performance of the package and compares the computer analyses with the actual test results.     

A Large Area Thin Nal(Tl) Detector for External Counting of Low-Energy Emitters in Lung

A new type lung monitor was designed for the detection of ²³⁹Pu and ²³⁵U in the lungs, and a preliminary calibration undertaken thereon. The detector consists of a 9.5 in. dia. by 0.5 in. thick NaI(Tl)-crystal and seven photo-tubes. Besides its use as an ordinary lung monitor, the instrument can be employed as an image detector giving a rough indication of the distribution of the activities in the lungs. This paper describes the structural arrangement of the detector, the preliminary calibration performed thereon and the limits of detection derived as function of the subject's effective tissue thickness.     

COSTA: A Computerized Support System for the Emergency Technical Advisory Body in Japan

If an accident resulting in an emergency occurs at a nuclear facility, the emergency technical advisory body (ETAB) is asked to make technical advice to the national government for emergency response. In order for the ETAB to carry out its role, a significant amount of materials containing information on nuclear power plants is prepared in the ETAB. These materials are all paper copies and very voluminous. Therefore it is required that they should be computerized for more effective use. Further, methods should be developed which aid the ETAB to understand the status of the affected facility and to predict potential radioactivity release to the environment.

To support the ETAB's activities in an emergency, the materials have been computerized as databases for more effective and quick use of them, and an expert system has been developed to quickly predict accident progression and environmental fission product release with the multivariate analysis method. The databases and the expert system have been unified to be a system of COSTA. The first phase of developing COSTA has been completed in FY1992. This paper describes the present status of development of COSTA.     

A method for enabling real-time structural deformation in remote handling control system by utilizing offline simulation results and 3D model morphing

A full scale physical test facility, DTP2 (Divertor Test Platform 2) has been established in Finland for demonstrating and refining the Remote Handling (RH) equipment designs for ITER. The first prototype RH equipment at DTP2 is the Cassette Multifunctional Mover (CMM) equipped with Second Cassette End Effector (SCEE) delivered to DTP2 in October 2008. The purpose is to prove that CMM/SCEE prototype can be used successfully for the 2nd cassette RH operations. At the end of F4E grant “DTP2 test facility operation and upgrade preparation”, the RH operations of the 2nd cassette were successfully demonstrated to the representatives of Fusion For Energy (F4E).Due to its design, the CMM/SCEE robot has relatively large mechanical flexibilities when the robot carries the nine-ton-weighting 2nd Cassette on the 3.6-m long lever. This leads into a poor absolute accuracy and into the situation where the 3D model, which is used in the control system, does not reflect the actual deformed state of the CMM/SCEE robot. To improve the accuracy, the new method has been developed in order to handle the flexibilities within the control system's virtual environment. The effect of the load on the CMM/SCEE has been measured and minimized in the load compensation model, which is implemented in the control system software. The proposed method accounts for the structural deformations of the robot in the control system through the 3D model morphing by utilizing the finite element method (FEM) analysis for morph targets. This resulted in a considerable improvement of the CMM/SCEE absolute accuracy and the adequacy of the 3D model, which is crucially important in the RH applications, where the visual information of the controlled device in the surrounding environment is limited.     

Scaling analysis for the OSU AP600 test facility (APEX)

In this paper, the authors summarize the key aspects of a state-of-the-art scaling analysis (Reyes et al., 1995. Westinghouse Electric Corporation, WCAP-14270) performed to establish the facility design and test conditions for the Advanced Plant Experiment (APEX) at Oregon State University (OSU). This scaling analysis represents the first, and most comprehensive, application of the Hierarchical Two-Tiered Scaling (H2TS) Methodology (Zuber, 1991. US Nuclear Regulatory Commission, Washington DC, NUREG/CR-5809) in the design of an integral system test facility. The APEX test facility, designed and constructed on the basis of this scaling analysis, is the most accurate geometric representation of a Westinghouse AP600 nuclear steam supply system. The OSU APEX test facility has served to develop an essential component of the integral system database used to assess the AP600 thermal hydraulic safety analysis computer codes.     

Design and analysis of a thermal hydraulic integral test facility for Bushehr nuclear power plant

In this paper, design and analysis of a thermal hydraulic integral test facility for Bushehr Nuclear Power Plant (NPP) is presented. The Bushehr Integral Test Facility (BITF) is a test facility designed to model the thermal-hydraulic behaviours of the Bushehr NPP (VVER-1000) pressurized water reactors currently in use in IRAN. These reactors have unique features that differ from other PWR designs. The BITF simulates the major components and systems of the reference NPP, making it possible to examine postulated small and medium break a loss of coolant accidents (LOCAs) and operational transients. The BITF is a volume-scaled model (1:1375). To ensure that gravitational forces remain equal to those in the reference reactor, the major components and systems in the BITF preserve 1:1 elevation equivalence to the reference reactor. The facility has four loops (each one consists of a hot leg, a steam generator, a loop seal, a main circulation pump and a cold leg), a pressurizer connected via two surge line to the hot leg of the loops 2, 4, the emergency-core-cooling system (ECCS) which is provided by an active pump simulating high and low pressure injection systems, and four hydro-accumulators. The report also contains a comparison between experimental data of PSB test facility and RELAP5 calculations of BITF facility under steady state condition of the reactor power 15% from the nominal.     

上海光源数字化束流位置信号处理器测试平台研制

数字化束流位置信号处理器(Digital Beam Position Monitor Processor,DBPM)是上海光源自主研发作为束流位置监测(Beam Position Monitor,BPM)的主要设备。针对DBPM电子学处理器产品化测试需求开发了一套基于MATLAB和程控技术的自动化测试平台,用于处理器批量测试验收和性能评估。平台使用Agilent MXG信号源和一分四功分器模拟4路真实探头信号,经过处理器采集处理后被上位机接收。测试仪器的设置和实验数据的获取,整个系统是基于客户端和服务器架构实现的。采用物理实验与工业控制系统(Experimental Physics and Industrial Control System,EPICS)的通道访问Lab CA接口实现数据的实时采集。目前该测试平台已经成功应用于大连自由电子激光装置和上海软X射线自由电子激光装置的DBPM处理器验收测试,处理器幅频响应曲线,不同衰减下模数变换器(Analog to Digital Converter,ADC)噪声水平测试、衰减步长一致性、通道间串扰及信噪比测试结果均达到设计指标。     

Disposal container safety assessment – drop tests with 'Yoyushindo-Disposal' waste container onto concrete target

Abstract

This paper presents technical details of the drop test performance as well as some experimental results of tests carried out with the Japanese 'Yoyushindo-Disposal' waste container for intermediate depth disposal. The drop test program comprised three single 8 m drop tests at the specimen's corner edge orientation onto a concrete slab. The slab was connected to the unyielding IAEA target of the BAM's 200 t drop test facility. The three tested specimens had masses between 20 000 and 28 000 kg depending on their content mass. The tests were accompanied by various metrology, such as strain and deceleration measurements, optical three-dimensional deformation methods, leak tightness testing and test installation for potential particle release measurements to collect a set of data for establishing a basis for safety assessment.     

An integral effect test of a complete loss of reactor coolant system flow rate for the SMART design using the VISTA-ITL facility and its simulation with the MARS-KS code

An integral effect test was successfully performed to provide data to assess the capability of the system analysis code to simulate a complete loss of reactor coolant system (RCS) flow rate (CLOF) scenario for the SMART (System-integrated Modular Advanced ReacTor) design. The steady-state conditions were achieved to satisfy initial test conditions presented in the test requirement, its boundary conditions were accurately simulated, and the CLOF scenario in the SMART design was reproduced properly using the VISTA-ITL facility. The natural circulation flow rate in the RCS was about 12.0% of the rated RCS flow rate and the flow rate in the passive residual heat removal system (PRHRS) loop was about 10.6% of its rated value in the early stage of the PRHRS operation. In this paper, the major experimental results of the CLOF test are discussed. The test results were analyzed using the best-estimate system analysis code, MARS-KS, to assess its capability to simulate a CLOF scenario for the SMART design.     

Study of startup transients and power ramping of natural circulation boiling systems

Numerical models of a natural circulation test facility and its prototype have been developed with RELAP5/MOD3.4 code and verified for their grid independence by nodal sensitivity studies. The model of the test facility has been validated for its steady state as well as transient predictions with the help of experimental observations. The transient predictions and parametric trends obtained by the numerical model of the prototype have been compared with those of the numerical model of the test facility. Thus, the ability of RELAP5 code to predict the transients during startup of a natural circulation boiling water reactor is verified. A powering procedure for the test facility has been conceptualized with the help of its RELAP5 model and demonstrated experimentally. Based on this, a similar powering procedure for the prototype has been proposed and simulated numerically with its RELAP5 model.