L. Di Pace

Results, conclusions, and implications of the SEAFP-2 programme

Abstract Fusion power stations inherently will have no actinides or fission products, extremely low levels of nuclear energy, and low levels of decay heat power. With appropriate design and material selection, these favourable inherent features could give rise to substantial safety and environmental advantages. Analyses performed within the SEAFP-2 project of the European fusion programme have shown that it should be possible to design commercial fusion power stations so that • the maximum doses to the public arising from the most severe conceivable accident driven by in-plant energies would be at the milliSievert level — well below the level at which evacuation would be considered; • after a few decades, most, perhaps all, of the activated material arising from the operation and decommissioning of the plant could be cleared or recycled, with little, or no, need for repository disposal; • the above goals can be achieved by using relatively well-developed and near-term low-activation martensitic steel as structural material. nThe results supporting these conclusions are summarised in this paper. The detailed lessons learnt will be input to a future European conceptual study of commercial fusion power stations.

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).

Materials-related issues in the safety and licensing of nuclear fusion facilities

Fusion power holds the promise of electricity production with a high degree of safety and low environmental impact. Favourable characteristics of fusion as an energy source provide the potential for this very good safety and environmental performance. But to fully realize the potential, attention must be paid in the design of a demonstration fusion power plant (DEMO) or a commercial power plant to minimize the radiological hazards. These hazards arise principally from the inventory of tritium and from materials that become activated by neutrons from the plasma. The confinement of these radioactive substances, and prevention of radiation exposure, are the primary goals of the safety approach for fusion, in order to minimize the potential for harm to personnel, the public, and the environment. The safety functions that are implemented in the design to achieve these goals are dependent on the performance of a range of materials. Degradation of the properties of materials can lead to challenges to key safety functions such as confinement. In this paper the principal types of material that have some role in safety are recalled. These either represent a potential source of hazard or contribute to the amelioration of hazards; in each case the related issues are reviewed. The resolution of these issues lead, in some instances, to requirements on materials specifications or to limits on their performance.

Cryostat pressurization in ITER during an ex-vessel loss of coolant accident sequence

Abstract The paper analyses the pressurization of the ITER cryostat during an ex-vessel loss of coolant accident (LOCA) in the first wall–shielding blanket (FW–SB) primary cooling loop. Since the cryostat is a strong barrier against the release of mobilized hazardous material, its integrity is essential in meeting the ITER safety objectives. The analyses refer to ITER TAC 4 design and were performed by means of a fast running thermal fluid dynamic code that allowed parametric studies. Helium LOCA can be handled by the code, so that the analyses did take into account phenomena such as coil quench and electrical arcing between coil turns, that may occur in tokamaks with superconducting cables in the magnet coils. The main result was that even in the worst conditions the pressure transient in the cryostat does not damage its integrity provided that the plasma facing components (PFC) cooling loops in the vacuum vessel remain undamaged. Hence, if no other energy sources than those coming from an ex-vessel LOCA are added to water, helium and heat structures in the cryostat, the cryostat maintains its integrity. More analyses are needed to evaluate an in-vessel LOCA and the consequent cryostat pressurization transient. In fact this event would add decay heat and exothermic chemical reaction as energy sources to be possibly taken into account in the pressurization transient. These analyses are not within the scope of this work.

Activated corrosion products in ITER first wall and shielding blanket heat transfer system

Corrosion and erosion phenomena play an important role in mobilising activated materials in fusion machines. This paper deals with the assessment of the activated corrosion products (ACPs) related to the primary heat transfer system (PHTS) of the first wall/shielding blanket (FW/SB) of the ITER plant. BCPs could be a cause for concern in terms of occupational radiation exposure (ORE) in maintenance scenarios. They could also be relevant in the case of severe accidents, such as ex-vessel LOCAs. The assessment mainly refers to the TAC-4 design developed for ITER. The mobilisation of the activated material has been estimated with the qualified CEA code PACTOLE. It considers ail the chemical and physical phenomena responsible for corrosion, activation and transport of corrosion products in cooling loops. The XSDNRPM-S code is used for neutronic calculations; the ANITA-2 code for activation calculations. The results obtained show the improvement gained, in terms of corrosion and radioactive inventory reduction, by avoiding the use of the boron as additive. Results obtained point out the impact of the main water chemistry parameters (e.g. water temperature and pH) on ACP production, transport and deposition. A parametric comparison has been carried out considering the coolant flowing during dwell periods, two different in-vessel FW/SB AISI 316L compositions and two fluences: 0.3 and 3 MW/spl middot/y/m/sup 2/.

Impact of plant incidents on worker radiation exposure for the SEAFP design

Fusion power reactors represent a significant source of energy supply with the potential for little adverse safety and environmental impact. Occupational radiation exposure, however, has been recognized as a potential area of concern, due to the significant quantities of activated products generated in a D-T fusion reactor and the need for frequent replacement of reactor components. This study was undertaken to provide a preliminary estimate of the potential impact of plant incidents on the annual station dose of two fusion reactor designs—SEAFP Model 1 (helium-cooled) and Model 2 (water-cooled). The assessment comprised an evaluation of the impact that plant incidents have on the annual station dose of fission reactor plants, and an assessment of the potential dose contribution of some incidents in a fusion reactor plant. The evaluation of water-cooled reactor data indicates that worker overexposures (20 mSv) from plant incidents are rare and that significant overexposures ( 50 mSv) are even less frequent—of the order of one in 600 reactor-years of operation. Furthermore, the number of workers exposed due to unplanned events is of the order of 0.2% of monitored workers. The evaluation of helium-cooled reactor data indicates better performance— no exposures above 20 mSv have been recorded. The assessment results for SEAFP Models 1 and 2 are consistent with the fission reactor results, however, only one category of incidents has been considered to date.

Environment source terms for ex-vessel FW/SB LOCA accident sequences in ITER EDA

The paper presents the environmental source term EST evaluation results for some ITER accident sequences, with reference to the activated materials contribution. The assessment is based on the end-of-1994 ITER baseline design and it refers to ex-vessel LOCAs in the first wall FW and shielding blanket SE heat transport system I-ITS. The main ITER characteristics are: fusion power 1.5 GW, average neutron power load on the outboard first wall 1 MW/m/sup 2/, fluence 3 MW-y/m/sup 2/, average pulse length 1000 s, dwell time 1200 s. The structural material for FW and SE is AISI 3161, (Mn 1.8 wt.%, Co 0.17 wt.%) stainless steel. Four independent loops for FW and SE heat transport system are considered. The European multi-code approach is briefly described jointly with the computer tools used for the assessment, A 1-D Sn neutron transport analysis in the cylindrical geometry has been considered for radioactive inventory RI evaluation by ANITA-2 activation codes. The mobilization of radioactive materials has been estimated with reference to the accident sequence phenomenology (e.g. the corrosion in HTS by using the PACTOLE code and the cooling blow down). The containment response to the accident sequences has been assessed (using the CONSEN code) to obtain the post accident thermal-hydraulic conditions inside the containment that are needed to estimate its retention factor. Finally, the NAUA code is considered to obtain the environmental source terms EST. A sensitivity analysis has been performed to verify the impact of the water chemistry on the environmental release composition and mass.

ORE Assessment Due to ACP in the PHTS of the Point Design Phase of the ITER Project

The work presented here dealt with the revision and the updating of the ORE (Occupational Radiation Exposure) assessment for the ITER PHTS (Primary Heat Transfer System). The data used come from the Point Design Documents and refers to the ITER design of the first half of 1996. The MCNP computer code was adopted to perform the shielding calculation. In addition, an accurate approach to evaluate the photon flux during maintenance and inspection activities was followed and recently published photon-flux-to-dose-rate conversion factors were applied to obtain the corresponding dose rate. The ACP inventory was taken from the relevant calculation performed with the PACTOLE code for the Point Design. A special ACP calculation was performed for each PHTS circuit and the related results are used in the respective dose rate calculations. The collective dose for the main activities performed to maintain the PHTS components is reported. The dose result for each activity type is shown and the comparison with a reference fission plant is discussed.

Radiological safety during maintenance of the primary heat transfer system of the ITER plant

The primary heat transfer system (PHTS) of the ITER plant is devoted to the heat removal from different plasma facing components: the first wall, the blanket, the divertor and the vacuum vessel. The system requires a scheduled or regular maintenance that involve component checking and changing, and a special maintenance that has a non periodic nature. During the maintenance operations some shielding barrier has to be partly or completely removed and workers are likely to be exposed to an unusual radiological dose. In the present work the occupational radiation exposure (ORE) for the maintenance activities performed at the PHTS has been assessed. The problem of the activated corrosion product (ACP) transport assessment inside the coolant has been faced and solved with the computer code PACTOLE. The working procedure protocols have been taken from the fission PWR plant experience. The shielding effectiveness of the PHTS components has been evaluated with a well tested computer code. The final result presented in the work is the collective dose for regular and special maintenance at the ITER PHTS, compared with typical data coming from the PWR operational experience.