Lessons learnt from ITER safety & licensing for DEMO and future nuclear fusion facilities

One of the strong motivations for pursuing the development of fusion energy is its potentially low environmental impact and very good safety performance. But this safety and environmental potential can only be fully realized by careful design choices. For DEMO and other fusion facilities that will require nuclear licensing, S&E objectives and criteria should be set at an early stage and taken into account when choosing basic design options and throughout the design process.Studies in recent decades of the safety of fusion power plant concepts give a useful basis on which to build the S&E approach and to assess the impact of design choices. The experience of licensing ITER is of particular value, even though there are some important differences between ITER and DEMO. The ITER project has developed a safety case, produced a preliminary safety report and had it examined by the French nuclear safety authorities, leading to the licence to construct the facility. The key technical issues that arose during this process are recalled, particularly those that may also have an impact on DEMO safety. These include issues related to postulated accident scenarios, environmental releases during operation, occupational radiation exposure, and radioactive waste.     

The impact of nuclear analysis on ITER design and licensing

The complexity of nuclear analysis on the design components of ITER is discussed. These analyses include the determination of several nuclear responses and it is shown that these results are not only relevant to the component under examination but have implications for the design of many other, often remote, systems. The contribution of nuclear analysis to the licensing process is discussed. An example is given of how this complexity means that the there is a large set of complementary analyses required to address the concerns of the licensing authorities. It also means that the nuclear analysis must be co-ordinated to ensure that the results are self-consistent and provide an integrated solution.     

Updated safety analysis of ITER

An updated version of the ITER Preliminary Safety Report has been produced and submitted to the licensing authorities. It is revised and expanded in response to requests from the authorities after their review of an earlier version in 2008, to reflect enhancements in ITER safety provisions through design changes, to incorporate new and improved safety analyses and to take into account other ITER design evolution. The updated analyses show that changes to the Tokamak cooling water system design have enhanced confinement and reduced potential radiological releases as well as removing decay heat with very high reliability. New and updated accident scenario analyses, together with fire and explosion risk analyses, have shown that design provisions are sufficient to minimize the likelihood of accidents and reduce potential consequences to a very low level. Taken together, the improvements provided a stronger demonstration of the very good safety performance of the ITER design.     

Numerical analysis of hydrogen risk mitigation measures for support of ITER licensing

In the ITER wet bypass scenario, water leakage, air ingress and hot dust (Be, W, and C) in the vacuum vessel could generate combustible hydrogen-air-steam mixture. Hydrogen combustion may threaten the integrity of the ITER VV and lead to radioactivity release. To prevent hydrogen energetic combustion, nitrogen injection system in VV and hydrogen recombination system in the pressure suppression tank (ST) were proposed. The main objectives of this analysis are to study the distribution of hydrogen-air-steam mixtures in the ITER sub-volumes, to investigate the feasibility of the nitrogen injection system to fully inert the atmosphere in the VV and to evaluate the capability and efficiency of the hydrogen recombination system to remove hydrogen in the ST. 3D computational fluid dynamics (CFD) code GASFLOW was used to calculate the evolution of the mixtures and to evaluate the hydrogen combustion risks in the ITER sub-volumes. The results indicate that the proposed hydrogen risk mitigation systems will generally prevent the risks of hydrogen detonation and fast deflagration. However, the atmosphere in ITER sub-volumes cannot be completely inerted at the early stage of the scenario. Slow deflagrations could still generate quasi-static pressures above 1 bar in the VV. The structural impact of the thermal and pressure loads generated by hydrogen combustions will be investigated in future studies.     

Tritium management and safety issues in ITER and DEMO breeding blankets

Safe, reliable and efficient tritium management in the breeder blanket faces unique technological challenges. Beside the tritium recovery efficiency in the tritium extraction and coolant purification systems, the tritium tracking accuracy between the inner and outer fuel cycle shall also be demonstrated. Furthermore, it is self-evident that safe handling and confinement of tritium need to be stringently assured to evolve fusion as a reliable technique. The present paper gives an overview of tritium management in breeder blankets. After a short introduction into the tritium fuel cycle and blanket basics, open tritium issues are discussed, thereby focusing on tritium extraction from blanket, coolant detritiation and tritium analytics and accountancy, necessary for accurate and reliable processing as well as for book-keeping.     

Preliminary tritium safety analysis on China DFLL-TBM for ITER

The dual-functional lithium-lead test blanket module (DFLL-TBM) system was proposed to be tested in ITER. A tritium permeation model of the entire DFLL-TBM system was developed, and the tritium permeation and inventory in DFLL-TBM system were done based on the model during normal operation. Three classes of off-normal situations had been preliminarily analyzed, i.e. in-vessel TBM coolant leaks, in-TBM breeder box coolant leaks and ex-vessel TBM ancillary coolant leaks. The results showed that some issues required significant R&D effort to guarantee the tritium release to the environment below the allowable level, such as the tritium extraction from LiPb and helium coolant and very efficient detritiation system. And more analyses would be carried in the future to further assess the safety of DFLL-TBM.     

Using the tritium plasma experiment to evaluate ITER PFC safety

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capabilty of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 × 10²³ ions/m².s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures.Prepared for the U.S. Department of Energy, Office of Energy Research under DOE Idaho Field Office Contract DE-AC07-76ID01570.     

Piping and component replacement in BWR systems; safety assessment and licensing decisions

For a successful operation of nuclear power plants it is important to demonstrate that major problems can be handled efficiently in terms of technical as well as regulatory actions to be taken. In the years 1973 to 1975 some cracks have been detected during the construction of piping systems in some German BWR plants. The materials used for the piping was 17 MnMoV 6 4 which is a precipitation hardening ferritic steel of a higher strength. To evaluate the safety implications of the problems encountered, a thorough reassessment of all BWR plants under construction or in operation has been performed by the responsible state licensing authority and the “Reactor-Safety Commission” on behalf of the Federal Ministry of the Interior. Furthermore the effects of cracks and degraded material conditions on the load carrying capability of the components were investigated by supplementary research programs. The problems have been solved by reinspection and repair and to the major part by the replacement of the piping and components affected. The replacement has been performed in a very successful manner on a narrow time-scale due to the close cooperation of all parties involved. The quality of the piping and components achieved resulted in a considerable improvement of the whole system. Secondary safety measures like pipe restraints which have some potential for a negative impact on flexibility and accessibility could be removed in cases where the license applied for it.     

Review of safety guidelines and licensing procedures applicable in Italy with regard to fusion machines

A consistent set of general safety criteria has been set up in Italy dealing with fusion machines, considering also the most recent recommendations issued by ICRP. The paper gives a short discussion of the more safety relevant aspects in the design of fusion machines starting from the consideration of the applicable dose limits. The procedure for the licensing of fusion machines is presented in the second part of the paper.     

Development of the PACTITER code and its application to safety analyses of ITER Primary Cooling Water System

The PACTITER code derives from the PACTOLE code, developed by the CEA for predicting activated corrosion products (ACPs) in PWR primary circuits. The operating conditions, material compositions and water chemistry of the various Primary Heat Transfer Systems (PHTS) of the International Thermonuclear Experimental Reactor (ITER) made mandatory the adaptation of the PACTOLE code.PACTITER was developed on the basis of dedicated experiments, namely devoted to determine copper solubility and stainless steel release in the ITER primary cooling systems conditions, which are rather different from those in PWR (i.e. water chemistry and temperatures). The PACTITER code has been extensively used in support of the ITER Generic Site Safety Report (GSSR) in the field of accident analysis and worker collective dose assessment.     

Safety and licensing of MHTGR

The Modular High-Temperature Gas-Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant's exclusion area boundary (EAB). The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error.     

ITER safety studies: The effect of two simultaneous perturbations during a loss of plasma control transient

The loss of plasma control events in ITER are safety cases investigated to give an upper bound of the worse effects foreseeable from a total failure of the plasma control function. Conservative analyses based on simple 0D models for plasma balance equations and 1D models for wall heat transfer are used to determine the effects of such transients on wall integrity from a thermal point of view.In this contribution, progress in a “two simultaneous perturbations over plasma” approach to the analysis of the loss of plasma control transients in ITER is presented. The effect of variation in confinement time is now considered, and the consequences of this variation are shown over a n–T diagram. The study has been done with the aid of AINA 3.0 code. This code implements the same 0D plasma-1D wall scheme used in previous LOPC studies.The rationale of this study is that, once the occurrence of a loss of plasma transient has been assumed, and due to the uncertainties in plasma physics, it does not seem so unlikely to assume the possibility of finding a new confinement mode during the transient.The cases selected are intended to answer to the question “what would happen if an unexpected change in plasma confinement conditions takes place during a loss of plasma control transient due to a simultaneous malfunction of heating, or fuelling systems?”Even taking into account the simple models used and the uncertainties in plasma physics and design data, the results obtained show that the methodology used in previous analyses could probably be improved from the point of view of safety.     

ITER operational availability and fluence objectives

The ITER availability objective is to reach for the machine operation in H phase an inherent availability of 60% and an operational availability of 32% assuming three 8 h plasma shifts operating mode and typically 8-month major shutdown after each 16-month experimental campaign. A functional analysis of the overall ITER machine from highest level functions down to main operational functions has been developed. The inherent availability (A_I) objective of ITER has been defined on the basis of a bottom-up approach and using the results of reliability, availability maintainability and inspectability (RAMI) analyses. The ITER strategy in terms of operational availability (A_O), Plasma pulse availability (A_P) and fluence objectives is not only to improve reliability by optimizing the design but also to gain the maximum of operation time by decreasing the scheduled downtime for preventive maintenance and increasing the maintainability of the operational functions, thus decreasing the frequency and the time to maintain or/and to repair.     

Safety and Environmental Activities for ITER

This paper will summarize highlights of the safety approach and discuss the ITER EDA safety activities. The ITER safety approach is driven by three major objectives: (1) Enhancement or improvement of fusion's intrinsic safety characteristics to the maximum extent feasible, which includes a minimization of the dependence on dedicated safety systems; (2) Selection of conservative design parameters and development of a robust design to accommodate uncertainties in plasma physics as well as the lack of operational experience and data; and (3) Integration of engineered mitigation systems to enhance the safety assurance against potentially hazardous inventories in the device by deploying well-established nuclear safety approaches and methodologies tailored as appropriate for ITER.     

ITER Vacuum Vessel design and construction

After implementing a few design modifications (referred to as the “Modified Reference Design”) in 2009, the Vacuum Vessel (VV) design had been stabilized. The VV design is being finalized, including interface components such as support rails and feedthroughs for the in-vessel coils. It is necessary to make adjustments to the locations of the blanket supports and manifolds to accommodate design modifications to the in-vessel coils. The VV support design is also being finalized considering a structural simplification. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. The detailed layout of ferritic steel plates and borated steel plates was optimized based on the toroidal field ripple analysis. A dynamic test on the inter-modular key to support the blanket modules was performed to measure the dynamic amplification factor (DAF). An R&D program has started to select and qualify the welding and cutting processes for the port flange lip seal. The ITER VV material 316 L(N) IG was already qualified and the Modified Reference Design was approved by the Agreed Notified Body (ANB) in accordance with the Nuclear Pressure Equipment Order procedure.     

The SLOWPOKE licensing model

The SLOWPOKE Energy System (SES-10) is a 10 MW heating reactor that has been developed in Canada. It will be capable of running without a licensed operator in continuous attendance, and will be sited in urban areas. It has forgiving safety characteristics, including transient time-scales of the order of hours. A process called “up-front” licensing has been evolved in Canada to identify, and resolve, regulatory concerns early in the process. Because of the potential market in Hungary for nuclear district heating, a licensing plan has been developed that incorporates Canadian licensing experience, identifies specific Hungarian requirements, and reduces the risk of licensing delays by seeking agreement of all parties at an early stage in the program.     

ITER vertical stabilization system

As part of the ITER design review, a reassessment of the specifications underlying the design of the vertical stabilization system (VS) was performed.Recent results from experiments, aimed at the evaluation of the feasibility of the ITER reference scenarios, have raised several concerns regarding mostly the ramp-up and ramp-down phases of the pulse. The main issue is the value of the internal inductance li which may reach values outside the range 0.7–1, considered as reference for the ITER control system design. Similar concerns apply to the low current L-mode plasmas, needed to the exploitation of the machine towards the development of the 15 MA pulse. The performance of the reference vertical stabilization system, under the revised conditions may be marginal, in particular if the effect of plasma generated noise on the velocity measurement is considered.A reliable and robust VS is mandatory to guarantee the operation of ITER at the reference elongation and plasma current values. To avoid de-scoping of the machine mission, several solutions have been proposed to improve the VS performances, ranging from an upgrade of the maximum voltage available to the present external coils system, to the introduction of in-vessel passive and/or active conductors.The paper presents an overview of the modelling and experimental effort aimed at the assessment of the baseline ITER VS and analyses the proposed solutions to improve the system performance.