[en] In the frame of the ITER Task 'Reference Accident Sequences', two accident sequences have been assessed to demonstrate the effectiveness of the use of integrated safety analysis code system (ISAS). The first one is a loss of coolant event in the divertor primary heat transfer system (DV PHTS) towards the vacuum vessel containment during normal plasma burn; the second one is a loss of coolant event in the DV PHTS towards the lower vault during baking conditions. The comparison of the results obtained in standalone and coupled modes was performed. The codes used for the analyses are ATHENA for thermal-hydraulic simulation, INTRA for containment studies and NAUA for aerosol transportation. The three codes are linked together in the ISAS chain by means of files written in the Gibiane language. The ISAS application seems to be promising in order to have a powerful tool in the fusion plant accident analysis. It is able to treat different codes, developed for different fields of applications, as a unique code simulating the complete evolution of an accident sequence. ISAS permits a detailed exchange of data between the codes, otherwise impossible in standalone mode. It reduces the errors in typing exchange data and takes into account the feedback effects during the calculation. The coupling of the codes makes possible to compensate the lack of some phenomenon models in one code if they are treated in another code of the chain (i.e. radiative heat exchange between heat structures that is missing in ATHENA and foreseen in INTRA). The results obtained in the two accident analyses demonstrate that ISAS can be very useful to represent in a more coherent way the trend of an accident respecting the correct balance of the inventories. In the paper the whole evolution of a typical fusion power plant accident is analysed starting from the break in the cooling loop system, studying the effects of the loss of coolant in the containment and simulating the leakage of the aerosols outside vacuum vessel and vaults

[en] The paper presents the activation analyses on Inconel-600 nickel-chromium alloy. Three activation data libraries, namely the EAF-4.1, the EAF-97 and the FENDL/A-2, and the FENDL/D-2 decay data library, have been used to perform the calculation with the European activation code ANITA-4/M. The neutron flux distribution into the material samples was provided by JAERI as results of 3D Monte-Carlo MCNP transport code experiment simulation. A comparison with integral decay heat measurement performed at the Fusion Neutronics Source (FNS), JAERI, Tokai, Japan, is used to validate the computational approach. The calculation results are given and discussed. The impact of the material composition, including impurities, on the decay heat of samples irradiated in fusion-like neutron spectra is assessed and discussed. The discrepancies calculations-experiments are within the experimental errors, that is between 6% and 10%, except for the short cooling times (less than 40 min after the end of irradiation). To improve calculation consistency with the experimental results, the knowledge of the material impurities content is mandatory

[en] The objective of the activity was to collect and analyse data coming out from operating experiences gained in the Joint European Torus (JET) for the Ion Cyclotron Resonance Heating (ICRH) system in order to enrich the data collection on failures of components used in fusion facilities. Alarms/Failures and malfunctions occurred in the years of operations from March 1996 to November 2005, including information on failure modes and, where possible, causes of the failures, have been identified. Beyond information on failures and alarms events, also data related to crowbar events have been collected. About 3400 events classified as alarms or failures related to specific components or sub-systems were identified by analysing the 25 hand-written logbooks made available by the ICRH operation staff. Information about the JET pulses in which the ICRH system was operated has been extracted from the tick sheets covering the whole considered time interval. 20 hand written tick sheets cover the period from March 1996 to middle May 2003, while tick sheets recorded as excel files cover the period from May 2003 to November 2005. By analysing the tick sheets it results that the ICRH was operated during about 12000 plasma pulses. Main statistical values, such as rates of alarms/failures and corresponding standard errors and confidence intervals, have been estimated. Failure rates of systems and components have been evaluated both with regard to the ICRH operation pulses and operating days (days in which at least one ICRH module was requested to operate). Failure probabilities on demand have been evaluated with regard to number of pulses operated. Some of the results are the following: - The highest number of alarms/failures (1243) appears to be related to Erratic /No-output of the Instrumentation and Control (I and C) apparatus, followed by faults (829) of the Tetrode circuits, by faults (466) of the High Voltage Power Supply system and by faults (428) of the Tuning elements. - The maximum number of events related to I and C (595) led to anomalous operations of CODAS, followed by 125 anomalous operations of stubs. - The total number of operation pulses for the four ICRH modules is of 44216; that corresponds to a total (integrated for the four modules) of 5280 days of pulse operation. - The number of failures/alarms of the ICRH system increases quite linearly with the number of pulses in which the system is operated. - A crowbar event happened on average every 9 ICRH pulses. - The rate of failure on demand of ICRH module is of about 0.10/pulse. (orig.)

[en] In the frame of the Generic Site Safety Report (GSSR) for the ITER experimental plant, several accident analyses have been carried out to analyse in detail the radiological risk linked with the possible releases. An ex-vessel loss of coolant followed by an in-vessel break, caused by the thermal stress on the divertor (DV) vertical target consequent to the plasma disruption, is the evolution of the accident referred in the study. The plasma disruption happens for the intervention of the fast plasma shutdown system occurring when the flow rate in the pump lowers below 80% of the nominal flow. This accident scenario is critical from the point of view of the vault pressurisation, of the activated materials mobilisation from the plasma chamber and of the hydrogen production inside the vacuum vessel (VV). The scope of the analyses presented in this paper is to quantify the influence of the ex-vessel break position on the accident consequences. A parametric study on the ex-vessel break position has been carried out. The pipe break before the heat exchanger is the most critical for the vault pressurisation and for the releases of activated corrosion products (ACP) and tungsten dust, while the break at the pump outlet maximises the hydrogen production inside the VV. The less critical accident is the break at the pump inlet for all the possible consequences

[en] This paper presents the ANITA-IEAF code package for the activation characterisation of materials exposed to neutrons with energies up to 150 MeV. It computes the radioactive inventories of materials exposed to neutron irradiation, continuous or stepwise. The activity, isotopic nuclide density, decay heat, biological hazard, clearance index and gamma ray source spectra are calculated at shutdown and at different cooling times. The code package is provided with a complete database that includes neutron activation data library, decay, hazard and clearance data library, and gamma library. The paper also presents an application of the ANITA-IEAF code package to the neutron exposure characterisation for the AISI 316 liner of the Test Cell area of the International Fusion Materials Irradiation Facility

[en] This paper presents the results of Sn radiation transport and activation calculations related to the ITER RC/RTO EU-I design, performed in support of safety and waste management analyses. The activation characteristics (included the clearance levels) have been estimated for the different materials/zones of the equatorial plane up to 10⁵ years after plasma operations. The Bonami-XSDNRPM sequence of the Scale 4.4 code system (using Vitamin-ENEA library, based on ENDF/B-VI data) has been used for radiation transport analyses. The ANITA-4M activation code (with FENDL/A-2 and FENDL/D-2 activation and decay data libraries) is used for the activation calculation. The unconditional clearance level data library is based on IAEA-TECDOC-855. First, a sensitivity analysis to optimise the radial spatial meshing for the neutron flux distribution evaluation and, accordingly, for the activation calculation, has been performed. Then, the clearance indexes of vessel and ex-vessel zones/materials have been calculated. The results are presented and discussed. A design option that considers copper instead of superconductor material for TFC winding pack has also been considered and assessed

[en] This study presents a method to obtain corrected self-shielded radiative capture cross-sections for tungsten isotopes to be used for activation calculations. The approach used is based on the application of the Bondarenko shielding factor method to the 175-group AMPX master library by means of the Bonami-Nitawl scale-4.3 sequence calculation. The ANITA-4M activation code calculates the tungsten radioisotopes production and the decay heat using the self-shielded cross-sections from ENDF/B-VI, JEF-2.2 and JENDL-3.2 data files. Two irradiation scenarios (5 min and 7 h) in the international thermonuclear experimental reactor (ITER)-like neutron flux spectrum defined by the fusion neutron source experiments are analyzed. The unshielded calculations result in discrepancy with experiment up to 70%, while the self-shielding treatment reduces drastically that discrepancy to less than few percents. In comparison to the experimental integral decay heat values provides a validation of the method used to deal with the self-shielding treatment

[en] The results of biological dose rate calculations for different locations in the vicinity of the concrete shield wall of the cryostat pit of the SEAFP-2 fusion plant are given in this paper. The Plant Model 3 (PM3) design (low-activation martensitic steel for first wall and blanket, Li₄SiO₄ as breeder material, helium cooled) is assessed and the results are compared with the corresponding ones related to the Plant Model 2 (PM2) design (low-activation martensitic steel for first wall and blanket, Li₁₇Pb₈₃ as breeder/multiplier material and water as coolant). Dose rates are evaluated both during plant normal running operation and after plasma shutdown before long-term maintenance activity. Contribution due to neutrons and gamma-rays are considered through a coupled n-γ Sn transport calculation. Effects of streaming are also evaluated. The dose rate values into pit cells result to be slightly higher for PM3 with respect to the PM2. In any case that area can be classified as 'Green zone', i.e. with unlimited access for radiation workers, according to the ITER zone classification

[en] A hypothetical double-ended pipe rupture in the ex-vessel section of the International Thermonuclear Experimental Reactor (ITER) divertor primary heat transfer system during pulse operation has been assessed using the nuclear source term ECART code. That code was originally designed and validated for traditional nuclear power plant safety analyses, and has been internationally recognized as a relevant nuclear source term codes for nuclear fission plants. It permits the simulation of chemical reactions and transport of radioactive gases and aerosols under two-phase flow transients in generic flow systems, using a built-in thermal-hydraulic model. A comparison with the results given in ITER Generic Site Safety Report, obtained using a thermal-hydraulic system code (ATHENA), a containment code (INTRA) and an aerosol transportation code (NAUA), in a sequential way, is also presented and discussed

[en] This paper focuses on some of the response functions obtained from a Sn radiation transport and activation analysis, namely the nuclear heating, decay heat and clearance index due to the neutron irradiation in the International Thermonuclear Experimental Reactor (ITER) fusion machine. The neutron and gamma flux spectra were calculated using the Scale 4.4a transport sequence Bonami-Nitawl-Xsdnrpm with a new 175n-42γ-coupled library (Vitenea-J) based on FENDL/E-2 data. The neutron and gamma heat deposition were evaluated using Kerma factor libraries based on EFF-2.4 data. The EASY-99 and, for quality assurance purpose, the ANITA-2000 code packages were used to obtain the activation characteristics of all the materials/zones of ITER. They include the specific activity, decay heat, contact dose, clearance index, list of isotopes at shutdown and dominant isotopes versus cooling time, related to each material. A total neutron fluence of 0.5 MW-y/m² at the outboard equator was considered. All the radiation transport and the EASY-99 activation results were provided to ITER Joint Central Team and they were used for the ITER Generic Site Safety Report (GSSR). This paper shows the relevant heat deposition results obtained from the radiation transport analysis and the activation characteristics (decay heat and clearance index) calculated with ANITA-2000. A comparison with the EASY-99 results is also given. The discrepancies between the two activation codes are lower than 1%