D.G. Cepraga

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.

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.

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/.

Cemented Containers Radiological Data From a Disused Uranium Mine Low-Level Waste Repository: A Calculated-Experiment Cross-Check for Data Verification and Validation

Code validation problems involve calculation of experiments and a comparison experiment-calculation. Experimental data and physical properties of these systems are used to determine the range of applicability of the validation. Once a sequence-code of calculations has been validated, it has to be underlined that the comparison experimental-calculated results involving “complex systems” or “complex experimental measures” permits also a bi-lateral cross-check between the calculation scheme and the experimental procedures. The results of the testing and the validation effort related to the collection of information and measured data and the comparison between code results with experimental data coming from a “low-level waste” repository are presented in this paper. The Baita-Bihor repository, sited into former disused uranium mine in Transylvania, has been considered as the source of experimental data. The study was developed through the following steps: a) collection and processing of measured data (radioactivity content and dose rate), from the cemented containers of the Baita-Bihor repository; b) decay gamma source calculation by the ANITA-2000 code package (the input data for the calculations are the measured isotope activities for each container); c) decay gamma transport calculation by the SCALENEA-1 shielding Sn sequence approach (Nitawl-Xsdrnpm-Xsdose modules of the Scale 4.4a code system, using the Vitenea-J library, based on FENDL/E-2 data) to obtain dose rates on the surfaces and at various points outside the containers; d) comparison experimental-calculated dose rates, taking into account also the measurement uncertainties. The new version of the ANITA-2000 activation code package used makes possible to assess the behaviour of irradiated materials independently from the knowledge of the irradiation scenario but using only data on the isotope radioactive material composition. Radioactive waste disposed of at Baita Bihor repository consists of worn reactor parts, resins and filters, packing materials, mop heads, protective clothing, temporary floor coverings and tools, the sources normally generated during the day-to-day operation of research reactors, the remediation-treatment stations and the medicine and biological activities. The low and intermediate wastes are prepared for shipping and disposal in the treatment stations by confining them in a cement matrix inside 220 litre metallic drums. Each container consists of an iron cladding filled by concrete Portland. Radioisotope composition and radioactivity distributions inside the drum are measured. The gamma spectroscopy has been used for. The calibration technique was based on the assumption of a uniform distribution of the source activity in the drum and also of a uniform sample matrix. Dose rate measurements are done continuously, circularly, in the central plan on the surface of the drum and 1 m from the surface, in the air. A “stuffing factor” model has been adopted to simulate, for the calculation, the spatial distribution of the gamma sources in the concrete region. In order to guarantee a complete Quality Assurance for codes and procedures, a simulation of the radioactive containers to evaluate the dose rates was done also by using the Monte Carlo MCNP-4C code. Its calculation results are in a very good agreement with those obtained by the Sn approach (discrepancies are around 2%, using the spherical approximation).Copyright

Resonance shielding-factor cross-section processing technique validation based on tungsten decay heat experimental data

Abstract 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.

Radiation transport and nuclear induced heating analysis for ITER baseline design

This paper presents the methodology calculation and the results of heat deposition due to the neutron and gamma interactions with the materials of different ITER zones. One dimensional discrete ordinate technique is used to perform the coupled 174-n and 38-/spl gamma/ transport analysis to produce the neutron and gamma flux distributions. The heat deposition is then calculated by means of the KERMA (Kinetic Energy Release in MAterials) factors and the neutron and gamma nux distribution of the related regions. The radiation transport assessment is based on the new VITAMIN-ENEA libraries derived from ENDF/B-V, FENDL and EFF-2 nuclear data. The kerma factor libraries have been obtained by processing basic nuclear data from EFF-2.4. The assessment is focused on the heat deposited in cryostat regions, because it is responsible for ozone formation from the possible oxygen sources that could be present into those regions.

Activation of TZM and stainless steel divertor materials in the NET fusion machine

Abstract This paper presents the results of the activation and decay heat calculations for the divertor plate materials of the Next European Torus (NET). The basic option assessed enables molybdenum alloy TZM and AISI 316L as material for divertor cooling channels. Burn time, effective irradiation time history, and fluence dependence on activation, decay heat, and contact dose is assessed. Impact of the material impurity level on the radioactive inventory is also investigated. The ANITA code is used, with updated cross sections and decay data libraries based on EFF-2 and EAF-3 evaluation files. The flux-weighted spectrum provided by XSDRNPM or ANISN 1-D codes has been used. The real NET geometry was modelled with the 3-D MCNP Monte Carlo neutron transport code.

Material activation assessment for waste analysis of the EU design of RC/RTO ITER

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 10 5 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.

Dose rate outside cryostat of the SEAFP-2 fusion plant

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 assessed considers low-activation martensitic steel for first wall and blanket, Li17Pb83 as breeder/multiplier material and water as coolant. For many analyses in the frame of the SEAFP-2 programme, one-dimensional modelling of the geometry is adequate. So far an infinite-cylindrical model representing the radial build at the plant mid-plane from the central solenoid up to the cryostat concrete wall was used. Dose rates are evaluated both during plant normal running operation and after plasma shutdown before long term maintenance activity. Contribute due to neutrons and gamma-photons are considered through a coupled n– Sn transport calculation. Effects of streaming due to penetrations are also evaluated. Radioactive decay of materials is included. The radiation transport analysis has been performed with the Bonami-Nitawl-XSDNRPM calculation sequence of the Scale 4.4 code system, using the Vitamin (ENEA) library, based on ENDF/B-VI cross section data. The Anita-4M code (with the FENDL/A-2 activation data library and the FENDL/D-2 decay data library) has been used for activation calculation to evaluate the gamma sources due to material radioactive decay. The dose rates outside the cryostat concrete wall have been obtained with the XSDOSE code using ANSI standard neutron and gamma flux-to-dose-rate factors.

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.