Proposal of Integrated Test Facility for in-Vessel Thermofluid Safety of Fusion Reactors

A vacuum vessel (VV) of a tokamak fusion reactor like the International Thermonuclear Experimental Reactor (ITER) consists the first confinement barrier that includes the largest amount of radioactive materials such as tritium and activation products. The ingress of coolant event (ICE) is a design basis event in the ITER where water is used as the coolant. The loss of vacuum event (LOVA) is also considered as an independent design basis event. Based on the results of ICE and LOVA preliminary experiments, an integrated in-vessel thermofluid test is being planned and conceptual design of the facility is in progress. The main objectives of the integrated test are to investigate the consequences of possible interaction of the ICE and the LOVA and to validate the analytical model of thermofluid events in the VV of the fusion reactor. This paper introduces a conceptual design of the integrated test facility and a testing plan.     

Dust tracking techniques applied to the STARDUST facility: First results

An important issue related to future nuclear fusion reactors fueled with deuterium and tritium is the creation of large amounts of dust due to several mechanisms (disruptions, ELMs and VDEs). The dust size expected in nuclear fusion experiments (such as ITER) is in the order of microns (between 0.1 and 1000 μm). Almost the total amount of this dust remains in the vacuum vessel (VV). This radiological dust can re-suspend in case of LOVA (loss of vacuum accident) and these phenomena can cause explosions and serious damages to the health of the operators and to the integrity of the device. The authors have developed a facility, STARDUST, in order to reproduce the thermo fluid-dynamic conditions comparable to those expected inside the VV of the next generation of experiments such as ITER in case of LOVA. The dust used inside the STARDUST facility presents particle sizes and physical characteristics comparable with those that created inside the VV of nuclear fusion experiments. In this facility an experimental campaign has been conducted with the purpose of tracking the dust re-suspended at low pressurization rates (comparable to those expected in case of LOVA in ITER and suggested by the General Safety and Security Report ITER-GSSR) using a fast camera with a frame rate from 1000 to 10,000 images per second. The velocity fields of the mobilized dust are derived from the imaging of a two-dimensional slice of the flow illuminated by optically adapted laser beam. The aim of this work is to demonstrate the possibility of dust tracking by means of image processing with the objective of determining the velocity field values of dust re-suspended during a LOVA.     

Real-time measurement of hydrogen generation level during beryllium dust oxidation by steam depending on the dust arrangement geometry

The paper concentrates on the safety issues in the International Thermonuclear Experimental Reactor (ITER) and describes the experiment on the measurement of hydrogen generation rate in case of Ingress of Coolant Event (ICE)—leak inside the vacuum vessel during interaction between water and beryllium (Be) dust. The ICE situation in ITER was simulated in a facility; the active spectroscopy was used to define the hydrogen content by the dynamics of oxidant concentration at a sampling frequency up to 10 Hz. Hydrogen release in time at temperatures of 500-900 °C is investigated, and different versions of dust arrangement are considered, i.e. on the surface and in a slot between armoring tiles at different initial density. The obtained results are compared with the known experiments.     

非能动余热排出系统安全分析程序SAC-PREARS 的功能与应用

SAC-PREARS 是一个用于分析非能动RHRS稳态和瞬态安全特性的专用程序.通过实验验证的用于AC-600 非能动 RHRS安全分析的MISAP 程序,对SAC-PREARS程序进行了稳态计算验证.并应用SAC-PREARS程序对200 MW 核供热堆非能动RHRS稳态和瞬态热工水力特性进行了分析,得出了具有工程意义的结论.     

Application of PSA Methodology to Design Improvement of JAERI Passive Safety Reactor (JPSR)

A probabilistic safety assessment (PSA) technique was applied to the design of JAERI Passive Safety Reactor (JPSR). A PSA was performed to clarify safety features and identify vulnerabilities of the original design. Based on the PSA results and considering thermal-hydraulic analyses and experiments, the JPSR design was improved to enhance plant safety. The improved design was re-evaluated with the PSA. Initiating events selected in this study were: large-break LOCA, medium- and small-break LOCAs, SGTR, main steam line break, loss of offsite power, loss of feed water, and other transients. Fault tree analyses were used to evaluate the system unavailabilities. The total core damage frequency due to internal events was estimated to be less than 10^?7/RY. The contribution of high frequency non-LOCA events could be significantly reduced by the design modification. The dominant initiating event was the small break LOCA and the dominant sequence was the failure of residual heat removal system. The present study indicated that the improved JPSR design has sufficient safety margin and the PSA methodology is very effective to improve reactor safety systems in a conceptual design phase.     

Addressing the heavy liquid metal – Water interaction issue in LBE system

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.     

中国实验快堆一级概率安全评价--事件树的建立

基于前期对初因事件的确定和归集研究，根据初因事件组的特征及对各初因事件序列的初步分析，确定了中国实验快堆(CEFR)一级概率安全评价(PSA)报告所要建立的事件树数目、各事件树的题头事件、事件序列后果的分类等。最后，根据CEFR具体安全设计特征创建了完整的事件树，为后续事件序列的深入分析奠定了重要基础。     

A preliminary assessment of beryllium dust oxidation during a wet bypass accident in a fusion reactor

A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). This beryllium dust oxidation model accounts for the diffusion of steam into a beryllium dust layer, the oxidation of the dust particles inside this layer based on the beryllium–steam oxidation equations developed at the INL, and the effective thermal conductivity of this beryllium dust layer. This paper details this oxidation model and presents the results of the application of this model to a wet bypass accident scenario in the ITER device.     

LWR Fuel Safety Research with Particular Emphasis on RIA/LOCA and Other Conditions

Fuel safety research at Japan Atomic Energy Research Institute (JAERI) is reviewed on the major subjects including studies on fuel behavior under postulated Reactivity Initiated Accident (RIA), postulated Loss of Coolant Accident (LOCA) and normal operating conditions. Nuclear Safety Research Reactor (NSRR) at JAERI has been utilized extensively for the studies of fuel behavior under RIA conditions. For the studies of fuel rod and cladding behavior under LOCA conditions, outpile experiments were conducted. The work on this subject has been concluded. Pellet Cladding Interaction (PCI) has been major subject on fuel integrity study during normal operating conditions. Irradiation experiments at Halden Boiling Water Reactor (HBWR) as well as code development are described.     

ROSA/APGOO Testing: Facility Modifications and Initial Test Results

The Japan Atomic Energy Research Institute (JAERI) and the United States Nuclear Regulatory Commission (USNRC) are jointly conducting confirmatory, integral testing on the Westinghouse AP600 reactor transient responses by using the ROSA-V Large Scale Test Facility of JAERI. This facility, built originally to simulate conventional 4-loop pressurized water reactors (PWRs), has been modified by adding components specific to the AP600 design. The modified LSTF now provides a full-pressure, full-height, 1130.5 volumetrically-scaled simulation of AP600. Five loss-of-coolant accident (LOCA) experiments were performed by August 1994, simulating transients initiated by cold leg breaks, a Pressure Balance Line (PBL) break, and inadvertently open Automatic Depressurization System (ADS) valves. These experiments indicated adequate core cooling and decay heat removal performance of the AP600 passive safety components.     

Preliminary thermal hydraulic safety analysis of water-cooled ceramic breeder blanket for CFETR

The Chinese fusion engineering test reactor (CFETR) was expected to bridge from the international thermonuclear experimental reactor (ITER) to the demonstration fusion reactor (DEMO). The water-cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for CFETR. In this paper, preliminary thermal hydraulic safety analyses have been carried out using the system safety analysis code RELAP5 originally developed for light water fission reactors. The pulse operation and three typical loss of coolant accidents (LOCAs), namely, in-vessel LOCA, in-box LOCA, and ex-vessel LOCA, were simulated based on steady-state initialization. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature can meet the design criterion which preliminarily verifies the feasibility of the WCCB blanket from the safety point of view.     

Test facilities for the study of thermofluid accidents in fusion reactors

Two large-scale test facilities with capabilities for a wide range of safety related testing are described. These facilities, situated at Stern Laboratories Inc. are being used to provide experimental data for demonstration and safety code verification of the characteristics of flashing water jets in fission reactor containments. The facilities are ideally suited to carry out thermofluid fusion safety testing such as the Japan Atomic Energy Research Institute Ingress of Coolant Experiment, in which the effects of a cooling water jet discharging from a broken tube and striking the plasma facing surfaces in the vacuum vessel will be investigated.     

Overview of SAFER03 Assessment Studies

The safety research for BWRs has been positively done by the JAERI, Japanese BWR utilities and BWR vendors in this decade and has shown the important phenomena under BWR LOCA conditions. Based on these significant results, the SAFER03 computer code was jointly developed by Toshiba, Hitachi and General Electric. SAFER03 has been qualified against the BWR simulation test data obtained from TBL, ROSA-III and FIST-ABWR test facilities. The objectives of this study are to assess the predictive capability of SAFER03 code to simulate the significant LOCA phenomena and to catch key parameters during BWR LOCA. This paper summarized the results of these SAFER03 assessments and showed that SAFER03 could predict the realistic behavior of BWR LOCA with slight conservative peak cladding temperatures.     

应对核与辐射突发事件的研究

介绍了河南省杞县放射源卡源事件的情况,分析了事件发生的经验教训,就应对核与辐射突发事件的不足之处进行了研究,提出了加强应对核与辐射突发事件工作的建议。     

Exchange Flow Characteristics in a Tokamak Vacuum Vessel of Fusion Reactor Under the Loss-of-Vacuum Conditions

When a Tokamak vacuum vessel of fusion reactor is broken, buoyancy-driven exchange flows will take place through breaches after the inside pressure of the vacuum vessel (VV) becomes equal to the outside pressure. The exchange flow may bring a mixture of activated dusts and tritium from the inside of the VV to the outside through the breaches. Moreover, the exchange flow may remove decay heat from the plasma-facing components. A preliminary LOVA (Loss Of VAcuum event) apparatus was constructed to investigate quantitative heat transfer characteristics of the exchange flows through the breaches under the LOVA conditions. The results of this study, the relationship between Froude numbers and breach locations in the VV was determined and empirical correlations for the average Froude numbers were derived.     

Progress of IRSN R&D on ITER Safety Assessment

The French “Institut de Radioprotection et de S?reté Nucléaire” (IRSN), in support to the French “Autorité de S?reté Nucléaire”, is analysing the safety of ITER fusion installation on the basis of the ITER operator’s safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.     

Evaluation of Dynamic Loads Induced by Chugging

A dynamic load evaluation method has been proposed for chugging phenomena which are assumed to occur and produce relatively large amplitude pressure spikes in the pressure suppression pool of a BWR containment, in case of a postulated loss of coolant accident. The proposed method is based on the analysis code developed by the authors and on theseven vent full scale tests performed at Japan Atomic Energy Research Institute (JAERI CRT), considering random nature of chugging phenomena. The dynamic loads are obtained by applying the design source functions of impulsive nature to the vent pipe exists in each BWR containment analysis model. The design source functions are defined to produce dynamic pressures which reasonably envelope the design spectrum based on JAERI CRT data in frequency domain.

As an application example, the dynamic loads induced by chugging have been assessed based on the proposed method and on the reported JAERI CRT data from the view point of conservative load evaluation.

The applicability of the analysis code has also been confirmed, since the simulated dynamic pressures have shown features and magnitudes similar to those observed in JAERI CRT.     

Experimental Investigation of Neutrons Penetrating through Two-Dimensional Heterogeneous Structure Consisting of Graphite and Aluminum

An experimental study has been performed on neutron penetration in a two-dimensional heterogeneous structure consisting of graphite and aluminum. The results provide a basis for verification of the accuracy of two-dimensional shielding calculation codes. The experiments were performed at JRR-4 of JAERI and included measurements of the thermal neutron flux for five different arrangements of graphite and aluminum. Calculations were performed with a conventional procedure using the two-dimensional removal diffusion code, TRD-1. The calculated thermal neutron flux of TRD-1 agrees with the measured value within about 25%.     

Study on premixing phase of steam explosion at JAERI

Melt jet breakup and fragmentation has been studied in ALPHA program at JAERI. In the first two experiments of the MJB series, a jet of molten lead–bismuth eutectic alloy was released into a deep pool of saturated water. The steam generation rate was measured and correlated with the jet behavior observed by a high-speed camera. The jet breakup length and debris size distribution were also evaluated. In parallel with the experimental study, JASMINE code has been developed for the simulation of a steam explosion process. The models of melt jet breakup and the particle breakup in the code were assessed by analyzing FARO-L14 and ALPHA MJB experiments.