G. Cambi

Failure mode and effect analysis on ITER heat transfer systems

The complexity of the ITER (International Thermonuclear Experimental Reactor) plant and the inventories of radioactive materials involved in its operation require a systematic approach to perform detailed safety analyses during the various stages of the project in order to demonstrate compliance with the safety requirements. The failure mode and effect analysis (FMEA) methodology has been chosen to perform the safety analysis at system level for ITER. The main purposes of the work are: to identify important accident initiators, to find out the possible consequences for the plant deriving from component failures, identify individual possible causes, identify mitigating features and systems, classify accident initiators in postulated initiating events (PIEs), define the deterministic analyses which allow the possible accident sequences to be quantified, both in terms of expected frequency and radiological consequences, and consequently, to ascertain the fulfillment of ITER safety requirements. This paper summarises the FMEA performed for the heat transfer systems (HTSs).

Collection and analysis of component failure data from jet systems: neutral beam injectors and power supply

The objective of this ongoing activity is to develop a fusion specific component failure database with data coming from operating experiences gained in various fusion laboratories. The activity began in 2001 with the study of the Joint European Torus (JET) Vacuum and Active Gas Handling Systems. Two years later the neutral beam injectors (NBI) and the power supply (PS) systems were considered and since last year the ion cyclotron resonant heating system is under evaluation. The number of failures/malfunctions that have occurred during the years of operations, failure modes and, where possible, causes and consequences of the failures were identified, as well as the sets of components under analysis. Components were classified and counted in order to determine component counts by type of component, related total operating hours and related demands to operate (for components operating in an intermittent manner). Main reliability parameters (such as the failure rate and corresponding standard errors and confidence intervals) for the component types were also estimated. In this paper the NBI and PS results are presented.

Neutronics and activation calculation for ITER generic site safety report

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 � /42g-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/m 2 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%. # 2002 Elsevier Science B.V. All rights reserved.

Source terms due to the activated corrosion products in primary cooling loops of ITER

Abstract This paper deals with the source terms due to the escape of activated corrosion products from a first wall or shielding blanket primary cooling loop of the international thermonuclear experimental reactor (ITER) machine following a loss of coolant accident (LOCA). The assessment is based on the European multi-code methodological approach set up to estimate the environmental releases of the activated corrosion/erosion products involved in accident scenarios of a fusion machine. The radioactive inventories (RIs) associated with the activation products have been estimated via the anita inventory code, using updated cross-section and decay data libraries based on EAF-3 activation data. The process source terms (PSTs) relevant to the LOCAs are estimated by considering mechanisms leading to RI mobilization (e.g. corrosion/erosion). The impact of the main operating parameters of the primary cooling loops is assessed. The FUMO codes have been used to estimate the thermal hydraulic conditions inside the containment. The PST transport and deposition in the containment are evaluated by the naua code in order to assess the released fraction from the last containment (reactor building).

A First Approach to the Safety Analysis of a Tokamak Test Reactor by a System Study Methodology

The safety analysis and risk assessment of a Tokamak Test Reactor is approached by an iterative, probabilistic, system study methodology, jointly devoloped by ENEAa and CEAb. The first part of this...

ANITA-IEAF: a code package for performing fusion material transmutation and activation analysis induced by intermediate energy neutrons

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.

Ex-Vessel Break in ITER Divertor Cooling Loop Analysis with the ECART Code

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. # 2003 Elsevier Science B.V. All rights reserved.

Environmental Source terms during a few Reference Accident Sequences (RAS) of NET/ITER plant

This paper outlines the multi-code methodological approach foreseen to estimate the environmental releases of the activated corrosion/erosion products involved in accidents scenarios of a fusion machine. The radioactive releases associated with some Reference Accident Sequences RAS for the Next European Torus NET have been calculated, based on the following items: Radioactive Inventory RI, Process Source Term PST, and Environmental Source Term EST evaluation. The assessment concerns the NET Basic Performance Phase. Various materials (such as AISI 316L, graphite, beryllium, and boron carbide for the PFC, and water as coolant) and operating conditions (normal running and conditioning) are considered. Different codes have been used and compared. The final environmental releases have been obtained by using RELAP5/Mod.3, FUMO, NAUA-Mod.5 M, and ANITA codes.

Summary of NET plant probabilistic safety approach and results by means of ENEA fusion plant safety assessment (EFPSA)

This paper summarizes the probabilistic safety assessment for the main accident scenarios associated with failures originating in the In-Vessel Plant Area of the Next European Torus (NET). The assessment refers to the Basic Performance Phase of operation under normal running and conditioning. For the corresponding accident sequences, the values of the annual expected frequency and the seriousness of consequences expressed as early dose to the Most Exposed Individual (MEI) of the public are listed.

Integrated safety analysis code system (ISAS) application for accident sequence analyses

Abstract 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.