Off-normal Conditions Research Articles

HELOKA data acquisition and control system: Current development status

HELOKA experimental facility, aimed to test the test blanket modules (TBM) mock-ups and prototypes, is under construction at the Research Centre Karlsruhe. HELOKA is an “8-shape” helium loop (HL), which feeds the TBM test section with helium (1.4 kg/s, 300 °C, 8 MPa). The maximum temperature at the test section outlet is 500 °C. HELOKA data acquisition and control system (DACS) comprises the facility control system and instrumentation. In addition to DACS and working independently from it, the central interlock and safety system (CISS) ensures the facility machine protection and personnel safety, with specific interlock logic, acting upon off-normal events or conditions. The paper briefly describes HELOKA DACS and CISS and reports in detail on the current development status: the installation and commissioning of DACS stage 1 consisting of the supervisory control system (SCS) and the control and monitoring for the water cooling system (WCS) and the related power supply.

Thermal-fluid flow analysis and demonstration test of a spent fuel storage system

Thermal-fluid flow analysis and demonstration test were performed for a spent fuel storage system. The commercial computational fluid dynamics (CFD) code, FLUENT was used for the numerical analysis. Effective thermal conductivities of a spent fuel assembly and a fuel basket were derived to optimize a thermal analysis model. Also, a porous model, which can simplify a complex configuration of a fuel assembly, was used in the thermal analysis. Demonstration test were performed to verify the thermal analysis method and procedure using a half scaled-down model and an electrically heated dummy fuel. The numerical analysis results were compared with the experimental data. Thermal analyses of the storage system were carried out for normal and off-normal conditions by using the verified analysis method.

On Power System Frequency Control in Emergency Conditions

Frequency regulation in off-normal conditions has been an important problem in electric power system design/operation and is becoming much more significant today due to the increasing size, changing structure and complexity of interconnected power systems. Increasing economic pressures for power system efficiency and reliability have led to a requirement for maintaining power system frequency closer to nominal value. This paper presents a decentralized frequency control framework using a modified low-order frequency response model containing a proportional-integral(PI) controller. The proposed framework is suitable for near-normal and emergency operating conditions. An <TEX>$H_{\infty}$</TEX> control technique is applied to achieve optimal PI parameters, and an analytic approach is used to analyse the system frequency response for wide area operating conditions. Time-domain simulations with a multi-area power system example show that the simulated results agree with those predicted analytically.

Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

A New Design for JT-60SA Toroidal Field Coils Conductor and Joints

The upgrade of JT-60U to JT-60 Super Advanced (JT-60SA), a fully superconducting tokamak, will be performed in the framework of the Broader Approach (BA) agreement between Europe (EU) and Japan. In particular, the Toroidal Field (TF) system, which includes 18 coils, is foreseen to be procured by France, Italy and Germany. This work covers activities from design and manufacturing to shipping to Japan. The present paper is mainly devoted to the analyses that lead to the conductor design and to the technical specifications of the joints for the JT-60SA TF coils. The conductor geometry is described, which is derived from Cable-In-Conduit concept and adapted to the actual JT-60SA tokamak operating conditions, principally the ITER-like scenario. The reported simulations and calculations are particularly dealing with the stability analysis and the power deposition during normal and off-normal conditions (AC losses, nuclear heating). The final conductor solution was selected through a trade-off between scientific approach and industrial technical orientation. Besides, the TF system connections layout is shown, derived from the industrially assessed twin-box concept, together with the associated thermo-hydraulic calculations ensuring a proper temperature margin.

The status of the ITER CODAC conceptual design

CODAC orchestrates the activity of 80–120 Plant Systems in normal ITER operation using a set of dedicated networks. Interlock systems protect ITER from potentially damaging operating off-normal conditions, using a segregated network. Safety systems protect the personnel and the environment and the nuclear safety systems will be subject to licensing. The principal challenges to be met in the design and implementation of CODAC include: complexity, reliability, transparent access respecting security, a high experiment data rate and data volume since ITER is an experimental reactor, scientific exploitation from multiple participant team experiment sites and the long 35-year-period for construction and operation. Two non-technical challenges drive the CODAC design. Firstly all technical plant systems follow an “in-kind” procurement model, as opposed to a direct commercial contractual relationship with suppliers. Secondly, the period of procurement will, itself, be long compared with the evolution of technologies in the fields of control and data acquisition.This paper addresses the various phases from definition of procurement agreements through to on-site commissioning of the project.

Thermo-hydraulic analysis of the windowless target system

The target system, whose function is to supply an external neutron source to a subcritical core in order to sustain the neutron chain reaction, is the most critical part of an ADS being subject to severe thermo-mechanical loading and material damage due to accelerator protons and fission neutrons. In order to reduce the material damage and to increase the life of the target system a windowless option was chosen in the framework of the European PDS-XADS project as reference configuration for the experimental ADS cooled by lead–bismuth eutectic alloy. This document deals with the results of the thermo-hydraulic analysis performed with STAR-CD and RELAP5 codes to assess the behaviour of the windowless target system during off-normal operating conditions. It also reports a description of modifications properly implemented in the codes for studying this kind of plant. The windowless target system shows a satisfactory thermo-hydraulic behaviour for the analysed accidents, except for the loss of both pumps without proton beam shut-off and for the beam trips lasting more than 1 s.

HEAT REMOVAL TEST USING A HALF SCALE STORAGE CASK

Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A dry storage cask to safely store the spent nuclear fuel should be able to adequately emit the decay heat from the spent nuclear fuel. Therefore, heat removal tests using a half scale dry storage cask have been performed to estimate the heat transfer characteristics of a dry storage cask under normal, off-normal, and accident conditions. In the normal condition, the heat transfer rate to an ambient atmosphere by convective air through a passive heat removal system reached 83%. Accordingly, the passive heat removal system is designed well and works adequately. In the off-normal condition, the influence of a half blockage in the inlet on the temperature appears minimal. In the accident condition, the temperature rose for 12 hours after the accident, but the temperature rise steadied after 36 hours.

The ITER CODAC conceptual design

CODAC orchestrates the activity of 60–90 Plant Systems in normal ITER operation. Interlock Systems protect ITER from potentially damaging operating off-normal conditions. Safety Systems protect the personnel and the environment and will be subject to licensing. The principal challenges to be met in the design and implementation of CODAC include: complexity, reliability, transparent access respecting security, a high experiment data rate and data volume since ITER is an experimental reactor, scientific exploitation from multiple Participant Team Experiment Sites and the long 35-year period for construction and operation. Complexity is addressed by prescribing the communication interfaces to the Plant Systems and prescribing the technical implementation within the Plant Systems. Plant Systems export to CODAC all the information on their construction and operation as “self-description”. Complexity is also addressed by automating the operation of ITER and of the plasma, using a structured data description of “Operation Schedules” which encompass all non-manual control, including Plasma Control. Reliability is addressed by maximising code reuse and maximising the use of existing products thereby minimising in-house development. The design is hierarchical and modular in both hardware and software. The latter facilitates evolution of methods during the project lifetime. Guaranteeing security while maximising access is addressed by flow separation. Out-flowing data, including experimental signals and the status of ITER plant is risk-free. In-flowing commands and data originate from Experiment Sites. The Cadarache Experiment Site is equated with the Remote Experiment Sites and a rigorous “Operation Request Gatekeeper” is provided. The high data rates and data volumes are handled with high performance networks. Global Area Networks allow Participant Teams to access all CODAC data and applications. Scientific exploitation of ITER will remain a human as well as technical challenge, to find methods of effectively combining the experience and observations of a geographically distributed research team.

A dose rate experiment at JET for benchmarking the calculation direct one step method

Neutrons produced by D–D and D–T plasmas induce the activation of tokamak materials and components. The development of reliable methods to assess dose rate is a key issue for maintenance and operation of nuclear machines, in normal and off-normal conditions. In support of ITER design activity an innovative computational tool based upon MCNP-4C Monte Carlo code has been developed to predict the dose rate after shutdown: it is called direct one step method (D1S). In the D1S approach the decay gammas are coupled to neutrons as in the prompt case and are transported in a single step in the same run. To check the capability of D1S method to predict dose rate in a tokamak a dedicated benchmark experiment was proposed for the 2005–2006 experimental campaign of JET. Two irradiation positions were selected for the benchmark: one inner position close to vessel and not far from plasma, called the two upper irradiation end (IE-2), while the second position is just outside a vertical port in an external position (EX). Passive detectors are used for in-vessel measurements: the high sensitivity thermo-luminescent dosimeters (TLDs) GR-200A (natural LiF), while one active detector of Geiger–Mueller (GM) type is used for out of vessel dose rate measurement. Both TLDs and GM detectors were calibrated in a secondary gamma-ray standard facility in terms of air-kerma. In the present work, residual background and dose rate due to early phase of JET 2005–2006 campaign are compared with the same quantities calculated using D1S approach. The impact of geometrical model and materials impurities to the calculated dose rates is discussed as well. The activity has been performed in the frame of the European Technological Tasks for JET JW5-FT5.20, in collaboration with United Kingdom Atomic Energy Agency (UKAEA) and Forschungszentrum Karlsruhe (FZK).

Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.

The MEGAPIE 1 MW target in support to ADS development: status of R&D and design

The MEGAPIE project is aimed at designing, building and operating a liquid metal spallation neutron target as a key experiment on the road to an experimental accelerator driven system and to improve the neutron flux at the PSI spallation source. The design of the target system has been completed. The target configuration and the operating conditions have been defined and the expected performance assessed. A preliminary safety analysis has been performed considering normal, off-normal and accident conditions and a corresponding report has been submitted to the authorities for licensing. The experience gained up to now shows that MEGAPIE may well be the first liquid metal target to be irradiated under high power beam conditions.

Overview of HTTR design features

The Japan Atomic Energy Research Institute (JAERI) designed and constructed the high temperature engineering test reactor (HTTR) in order to establish and upgrade the technology basis for the high temperature gas-cooled reactor (HTGR) and develop the technology for high temperature heat applications. The HTTR is a helium-cooled and graphite-moderated HTGR with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C. The first criticality was attained on November 10, 1998 and the rated power operation with the reactor outlet coolant temperature of 850 °C was achieved on December 7, 2001. Since 2002, safety demonstration tests simulating anticipated operational occurrences such as decrease of primary coolant flow-rate and reactivity insertion have been carried out to ensure safety during off-normal conditions. From 2005, various irradiation tests for fuels and materials will start. In addition, a hydrogen production test facility will be coupled with the HTTR by 2015 to produce hydrogen by nuclear. The history and future plan, major design features and R&D programs of the HTTR are summarized in this paper.

Modelling and Control of a TiNi Shape Memory Alloy (SMA) Actuator

TiNi SMA is a good candidate for nanoactuation due to its low operation voltage, large range of recoverable deformations, and high recovery forces. This paper addresses the modelling and control of a TiNi SMA wire in free-standing configuration. The dynamic behaviour of the wire during the cooling-heating phase under the normal operation condition (2~3V) and the off-normal operation condition (1.2~1.8V or 3.2~3.4V) are studied. A first-order model is employed to characterize the system. A hybrid input shaping, consisting of normal and off-normal operation voltage, is developed to achieve the ideal wire position control. A vision system measures the displacement of the wire, and then sends the position information to the controller to adjust the input to achieve the ideal wire position.

Use of Thoria-Urania Fuels in PWRs: A General Review of a NERI Project to Assess Feasible Core Designs, Economics, Fabrication Methods, In-Pile Thermal/Mechanical Behavior, and Waste Form Characteristics

This paper provides an introduction to and a summary of the remaining papers in this issue of Nuclear Technology. The papers in this issue present the important results from a U.S. Department of Energy–sponsored Nuclear Engineering Research Initiative (NERI) project to study the efficacy of the thorium-uranium dioxide (ThO2-UO2) once-through fuel cycle in current light water reactors. The project addressed fuel cycle neutronics and economics; ThO2-UO2 fuel manufacturing; the in-pile thermal/mechanical behavior of ThO2-UO2 fuel during normal, off-normal, and accident conditions; and the long-term stability of ThO2-UO2 waste. Results from this work show that a small-scale separation of the uranium and thorium will enhance the fuel reactivity and achievable burnup from uranium-thorium dioxide fuels. Under conditions that meet the thermal requirements in present pressurized water reactors (PWRs), a properly designed microheterogeneous fuel will have more reactivity than all-uranium fuel, and the overall production of plutonium is significantly reduced. The use of thorium as a host for actinide fuels when PWRs are used for actinide transmutation was also explored. It was also determined that there were no fundamental obstacles to converting the current plants that manufacture uranium oxide–only fuel to a mixed ThO2-UO2 fuel. Also, the in-service and transient thermal and mechanical performance of homogeneous ThO2-UO2–based fuels with respect to safety is generally equal to or better than that of all-uranium fuel. Furthermore, a mixed thorium-uranium dioxide spent fuel appears to be a much more stable waste form than uranium oxide spent fuel.

A New Series-Connected Hybrid Magnet System for the National High Magnetic Field Laboratory

The National High Magnetic Field Laboratory (NHMFL) proposes to build a new hybrid magnet system with three specific goals: 1) address the needs of users requiring high homogeneity, high temporal stability, and long residence times at moderately high fields; 2) enhance user service through use of a single 10 MW power supply (allowing simultaneous operation of multiple systems); and 3) make the system attractive in terms of combined capital and operating costs. The basic configuration are series-connected hybrid, wherein resistive insert and superconducting outsert are connected in series and powered by a single 10-MW unit (20 kA and 500 V) of the NHMFL dc power system, a configuration with advantages for handling faults and off-normal operating conditions and with improved temporal stability. Specific design goals are 35 T, 10-ppm uniformity over a 10-mm DSV, and access for 40-mm diameter probes. By powering the superconducting outsert with 20-kA high-temperature-superconductor (HTS) current leads, cryogenic loads are kept small, substantially smaller even than the present 45-T Hybrid. The net result is a magnet design: 1) with capability for both higher field and higher field quality than present resistive-only systems, 2) that permits simultaneous service of two or more users, 3) that is sufficiently compact to fit within the standard resistive-magnet cell, 4) whose ease of operation is comparable to resistive-only systems, and 5) whose lifetime cost (including construction and operation) can be significantly lower than comparable resistive-only systems. We discuss details of the existing conceptual design, the methodology for its creation, the perceived development needs, and the projected program to realize this system.

Two-phase turbulent mixing and buoyancy drift in rod bundles

This paper describes the development of generalized relationships for single- and two-phase intersubchannel turbulent mixing in vertical and horizontal flows, and lateral buoyancy drift in horizontal flows. The relationships for turbulent mixing, together with a recommended one for void drift, have been implemented in a subchannel thermalhydraulics code, and assessed using a range of data on enthalpy migration in vertical steam–water flows under BWR and PWR diabatic conditions. The intent of this assessment was to optimize these relationships to give the best agreement with the enthalpy migration data for vertical flows. The optimized turbulent mixing relationships were then used as a basis to benchmark a proposed buoyancy drift model to give the best predictions of void and enthalpy migration data in horizontal flows typical of PHWR CANDU 1 CANDU=CANada Deuterium Uranium, registered trademark of Atomic Energy of Canada Ltd. 1 reactor operation under normal and off-normal conditions. Overall, the optimized turbulent mixing and buoyancy drift relationships have been found to predict the available data quite well, and generally better and more consistently than currently used models. This is expected to result in more accurate calculations of subchannel distributions of phasic flows, and hence, in improved predictions of critical heat flux (CHF).

Advanced tokamak operation using the DIII–D plasma control system

The principal focus of experimental operations in the DIII–D tokamak is the advanced tokamak (AT) [Proc. 18th Symp. on Fusion Technol., August, 1994, Karlsruhe, Germany, Elsevier, Amsterdam, 1 (1995) 731] regime, which requires highly integrated and flexible plasma control to achieve. In a high performance AT, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII–D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response.

Tritium Issues Toward Fusion Plants

ABSTRACTTechnical issues on tritium technology are investigated from the aspects of processing, safety and fuel supply, considering the concept of DEMO plant following ITER as the next target. Fusion plant equipped with power blanket will contain high temperature heat transfer medium and bred tritium in a tritium cycle. Although the inventory and throughput may not increase drastically from ITER, tritium plant will require significant technical improvements characterized by the self-consistent tritium fuel cycle and safety function to maintain the tritium level in the power train at adequately low level. Tritium balance issue will be one of the most important features because it will strongly affect the fusion in energy market by supplying initial loading. Tritium processing for coolant that will be mainly used for normal operation will dominate the safety feature of the entire plant by its technical difficulty and importance. Under off-normal conditions, this coolant tritium process can remove possible spill within the confinement, and thus fusion plant will not have any major process dedicated for accidents. Tritium technology is essential to make fusion energy attractive from the aspect of socio-economics, and its success in development is of vital importance toward fusion power plant as viable energy source for future.

Study on Sudden Loss of Cryogenic Coolant Accident Happened in the Hydrogen Isotope Separation System for Fusion Reactor

ABSTRACTSudden loss of cryogenic helium coolant accident (ISS-LOCA) in the cryogenic distillation columns for the hydrogen isotope separation system (ISS) is one of the worst situations because it leads the evaporation of liquid hydrogen in the column. From this background, an intended ISS-LOCA test was conducted with an actual ITER-scale cryogenic distillation column. Sudden increase of internal pressure was not observed and enough time is found to recover the hydrogen isotope into a storage system if vacuum insulation is maintained and reboiler heaters are turned off immediately. In off-normal conditions, the rapid recovery of hydrogen in the column by an empty hydrogen storage bed is a reasonable hydrogen recovery scenario. Validity of the hydrogen recovery scenario was proved by a demonstration test.