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Summary results of 2005 activation calculation in support of ITER. EFDA Task No. TW5-TSS-SEA4.1/D3, Final Report, ENEA FUS-TN-SA-SE-R-135
No full textDuring 2005 various analyses have been performed by ENEA in the frame of the EFDA task TW5-TSS-SEA4.1 in support of ITER licensing. Two aspects have been assessed (milestones M1 and M2, respectively):
M1: the first one is related to the impact of the cobalt content into the materials of the Vacuum Vessel, with respect to their clearance possibility and,
M2: the second one is related to the calculation of the activation of ITER coolant pipes that have been irradiated in OSIRIS fission reactor and then inserted into the CORELE-2 loop (in CEA centre of Cadarache) for material release rate tests and PACTITER code validation.
The results related to M1 and M2 were presented in details and discussed respectively in two ENEA reports (deliverables D1 and D2, respectively):
D1: ENEA FUS-TN-SA-SE-R-131, July 2005, plus Addendum (September 2005).
D2: ENEA FUS-TN-SA-SE-R-133, October 2005.
The relevant results obtained are summarised into this final report (Deliverable D3), together with updated analyses of the clearance possibility of ITER Vacuum Vessel materials when the new (August 2004) unconditional clearance levels given in IAEA Safety Guide RS-G-1.7 are used instead of those given in the older (January 1996) IAEA TECDOC-855. The relevant outcomes from that updated analysis are:
• the 430 ferritic steel and the SS 304B4 steel of the Outboard VV zone VVSHDO(2) result to be clearable after a longer time if compared with the previous analysis results when TECDOC-855 clearance level data were employed. This is particularly significant for the SS 304B4 steel that becomes clearable only after about 6000 years (with respect to the 90 years of the previous analysis).
• the remarkable change for the VVSHDO(2) – SS 304B4 steel is due to the highest contribute (at 100 years cooling time) from the Ni-63, which is now (i.e. with the RS-G-1.7) of about 40% with respect to the older (i.e. with the TECDOC-855) value of about 10%.
Detailed results of the new activation calculation are given in the appendix
Estimation of IFMIF Test Cell Shielding Performance by 3D Calculations (MCNP-TORT)
No full textIn order to contribute to establish shielding design criteria, the Bologna group has produced a package of libraries, codes and procedures, for multidimensional shielding IFMIF evaluations: Vitenea-IEF, a new intermediate energy coupled 256-neutron till 150 MeV and 49-gamma-ray till 100 MeV multi-group cross-section library in Ampx format, suitable for Sn radiation transport codes; ANITA-IEAF, a code package able to manage the many reaction channels that open for neutron energies higher than 20 MeV and up to 150 MeV; Libtort, an utility sequence for the production of working libraries for TORT; Proctort, an utility module able to manage the TORT output data file in order to retrieve, process and calculate the various nuclear responses, i.e. neutron and gamma fluxes and dose rates.
The 2003 ENEA activity was devoted to the qualification and application of 3D methods, TORT and MCNP-4C2, for the IFMIF dose rate evaluation. The MCMP calculations estimated the dose rates in the Test Cell and its surrounding areas for a “safety conservative geometry”, Concrete ANS, 405 cm, He- 20%, for Beam-on/Start of cycle operational phase, at various detector surfaces in different directions. The total dose rates for the Frontal, Lateral, and Floor directions are less than the recommended one (10μSv/h)(Access/Maintenance room, Test Cell Technology room and Temporally Storage Pits).
In order to check the shielding effectiveness, TORT calculations were performed considering various shielding configurations differing in: concrete material composition, He-cooling system treatment and Test Modules location. The Vitenea-IEF library was converted to TORT format via SCALENEA-1 (application of the Bondarenko shielding factor treatment) and GIP module. Two Test Modules locations were considered. The concrete JAERI heavy shows a better shielding performance than the Standard ANS one. The dose rates when the cooling is considered are about 3-4 times higher than when it is not. The presence of the Test Modules inside the Test Cell modifies 2-3 times only the dose rates at the frontal surface.
The MCNP-4C2, TORT and 1D SCALENEA-1 dose rate results for the “safety conservative geometry” are in good agreement, within 15% (Frontal direction, 2.55 μSv/h, 2.33 μSv/h and 2.45 μSv/h, respectively). The 1D spherical-thickness-equivalence-approach appears to be a suitable fast method to perform preliminary IFMIF dose rate calculations, useful as an auxiliary tool with 3D methods. The assessment performed with TORT has outlined that important parameters that characterise the shielding design are the concrete composition, its compactness (He-cooling treatment, etc.) and, especially for the Frontal direction, the location-identification of Test Modules inside the Test Cell. As a general conclusion, for shielding estimation and evaluation of all the radiological hazard sources in IFMIF facility, MCNP-4C2 is used as reference code system while TORT for detailed calculations, due to its friendly handling-processing input and output data and its less time consuming, hours instead of days, than Monte Carlo methods
ANITA-2007 Activation code package: main features and validation
No full textANITA-2007 is a code package for the activation characterisation of materials exposed to neutrons in fusion machines. The main component of the package is the activation code ANITA-4U that computes the radioactive inventory of a material exposed to neutron irradiation, continuous or stepwise. It provides activity, atomic density, decay heat, biological hazard, clearance index and gamma-ray source spectra at shut down and at different cooling times.
Main improvements with respect to the previous release of the activation code package (the ANITA-2000) are: the possibility to choose among updated neutron activation data libraries, the evaluation of the decay gamma source in the Vitamin-J 42- energy group structure and the introduction of updated clearance levels data (derived from the IAEA Safety Guide, No. RS-G-1.7).
Different 175-group neutron activation libraries (based on FENDL/A-2.0, EAF-2003, and EAF-2005.1 neutron activation cross section data) are included in ANITA-2007 code package.
The decay, hazard and clearance data library includes decay data based on the FENDL/D-2.0, hazard data based on the ICRP-72 committed effective dose equivalent from inhalation and ingestion (in Sv/Bq) of radio-nuclides and two sets of clearance levels data (the first set contains the “unconditional clearance levels”, based on IAEA-TECDOC-855 and the second one contains the updated values based on IAEA Safety Guide, No. RS-G-1.7).
The gamma library (in the 42- Vitamin-J energy group structure) is based on the FENDL/D-2.0 evaluated decay data file.
Up to 2000 irradiation time intervals can be considered and a different neutron loading can be used for each irradiation time interval.
The updated code package ANITA-2007 has been tested by running sample problems and comparing the results with those obtained with ANITA-2000 and with EASY-2005 code packages. No differences between ANITA-2007 and ANITA-2000 activation parameters calculations and discrepancies within 1% between ANITA-2007 and EASY-2005.1 results have been found.
The ANITA-2007 code package version running on a PC (Windows 98, 2000,XP) is actually distributed by ENEA. A CD-Rom containing the executable file ANITA-4U, all the necessary data libraries, the how-to-use manual and few test-case problems will be made available, after ENEA approval, upon request to:
Dr. Tonio Pinna, ENEA FUS/TEC - Fusion Technology, Safety & Environment Section, Via Enrico Fermi 45, I-00044, Frascati (Rome), Italy. Email: [email protected]
Activation calculations in support of PACTITER validation based on CORELE-2 loop measurements
No full textThe report presents the analyses done, in the frame of the EFDA Task No. TW5-TSS-SEA4.1 (milestone 2) during the year 2005 to support the PACTITER code validation needs
The results include the reaction rates and the isotope activities of the 316LN steel test tubes that were irradiated in OSIRIS reactor (CEA-Saclay) during January 2004 and then inserted in CORELE-2 loop (CEA-Cadarache) for ITER material corrosion tests and PACTITER code validation. PACTITER The report describes also the approach that has been used for.
A neutron model has been preliminarily set up to perform the radiation transport analysis with the SCALENEA-1 sequence and to identify (through a comparison with experimental neutron flux data from the OSIRIS team) the suitable neutron flux spectra for ANITA-2000 activation calculations.
A comparison of the ANITA-2000 activation calculation results with experimental activity measures performed on the irradiated test tubes and with calculation performed at CEA using the CASTOR code is also presented. That comparison indicates a good agreement between experimental data and calculations.
Moreover, because from years 2005 to 2007 new experimental corrosion measurements of the test tubes irradiated into the OSIRIS reactor are foreseen in CORELE-2 loop, some predictions on the expected tube material activity parameters up to the year 2008 are presented
The SCALENEA-1 multipurpose Sn calculation sequence for application in fusion field: main features and its validation based on experimental data from a low-level waste repository
No full textThe paper presents the main features of the Sn calculation sequence SCALENEA-1 that has been extensively used for radioactive inventories, source terms, wastes and dose rates calculation for ITER, SEAFP and IFMIF fusion machines.
The sequence includes the following steps:
1. nuclear data processing (from Evaluated Nuclear Data File), to produce master and working data libraries (problem dependent);
2. radiation (n, g) transport analysis, to produce neutron and gamma spectra;
3. radiation transport results post-processing, such as: a) radioactive inventories (activation calculations with activation code packages, e.g. the ANITA-2000), b) radiation damage, c) dose rates, and d) nuclear heating.
As part of the validation process for the code package and for the calculation sequence, experimental-calculation comparison has been performed using the information and the measured isotope radioactive inventories and dose rate gathered from drum containers of a “low-level waste repository”. 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 repository (data from 35 drums were analyzed); 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 (Bonami-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.
In order to guarantee a complete Quality Assurance for codes and calculation scheme, a simulation of the radioactive containers to evaluate the dose rates was done also by using the Monte Carlo MCNP-4c code.
The following conclusions can be outlined from the result analysis:
- Agreement with the (gamma dose rates) experimental data: discrepancies (C-E)/E lower than 50%.
- Sn (Scalenea-1) and Monte Carlo (MCNP 4c) comparison: discrepancies lower than 1%.
The analyses done provided a contribution to the Validation process of the ANITA-2000 activation code package and, more extensively, of the overall calculation sequence used to perform activation calculations for ITER GSSR.
As a general conclusion, the Scalenea-1 calculation sequence is able to provide reliable results for various ITER safety analyses purposes
Summary report on the 2004 ENEA 1D-Sn neutronics and activation calculations for ITER
No full textThe present report outlines the results of new contributes to the 1D-Sn neutronics and activation calculation performed by ENEA in support of ITER safety analyses. They have been done in the frame of the EFDA Task No. TW4-TSS-SEA4.1.
They were split in three parts:
a) Updating and maintenance of the tools, libraries and procedures for 1-D radiation transport analysis and activation calculations.
The new updated calculation approach SCALENEA-1 uses the VITENEA-J library, for radiation transport analyses, and both FISPACT-2003 or ANITA-2000, with EAF-2003 neutron activation library, for activation calculations.
b) Calculation in support of ITER Primary Heat Transfer System ACPs assessment for Occupational Radiation Exposure evaluation.
Updated Stainless Steel SS316 LN-IG inventories related to wet zones of ITER First Wall and Blanket coolant loops have been obtained in support of the activation characterization of the Activated Corrosion Products (ACPs). Moreover the impact of Cobalt content and of different irradiation scenarios on the coolant ACPs radioactive inventories were analysed.
c) Examples of updated ITER in-vessel material activation characterisation.
As applicative examples of the updated calculation approach, the activation characteristics of some ITER in-vessel outboard zones/materials have been obtained. The results obtained have been compared with those included into the Vol. III and V of GSSR.
The results presented in this report point out once again the high flexibility of the SCALENEA-1 calculation sequence. It represents a very useful support in performing activation calculation for different purposes in safety analyses, particularly when fusion machine design is not yet frozen. Moreover it is also useful during design phases to perform parametric studies
OSIRIS neutronic and activation simulation with Scalenea-ANITA in support of PACTITER/CORELE analyses
No full textThe present technical note describes the procedure that has been used to obtain the reaction rates and the activation data related to the test tubes irradiated into the OSIRIS reactor and inserted into the CORELE-2 loop for the corrosion/deposition experimental measurements
ANITA-2000 activation code packages: 2005 validation effort against Karlsruhe Isocyclotron and FNG-ENEA experiments. EFDA Task No. TW5-TSS-SEA5.5/D7, Final Report, ENEA FUS-TN-SA-SE-R-136
No full textThe report summarizes the results of new contributes to the validation effort of the activation code package ANITA-2000 performed in the frame of the EFDA Task No. TW5-TSS-SEA5.5-D7:
Two main validation exercises have been carried on and discussed in the report:
A. Comparison between vanadium alloys, nickel, copper, lithium orthosilicate, Eurofer-97 steel and tungsten samples specific activity measurements performed at the Karlsruhe Isochronous Cyclotron (KIZ) and ANITA-2000 estimates based on the EAF-2003 activation library.
B. Comparison between Molybdenum and Tantalum photon and electron decay heat measurements from FNG ENEA-Frascati and ANITA-2000 / EAF-2003 calculation results.
The comparison of ANITA-2000 activation code package calculation with experiments indicate generally a good agreement representing a further step in the code validation process.
As a general conclusion from the results analyses, it can be observed that the ANITA-2000 code package with the EAF-2003 neutron activation library is a good tool for activation estimates in fusion field environment
Evaluation of activation, inventories and dose rates induced by deuterons and neutrons along the RFQ and DTL parts of IFMIF Accelerator due to deuteron beam losses
No full textThe present report summarises the main features of the methodology approach developed for this assessment and its application to the IFMIF accelerator parts: the RFQ (Radio Frequency Quadrupole), that accelerates the deuteron beam up to 5 MeV, and the DTL (Drift Tube Linac) that accelerates the deuteron beam from 5 MeV up to 40 MeV.
The analysis points out that in the RFQ deuteron energy range the nuclear models used by MCNPX-2.5b code don’t produce secondary neutrons in the deuteron nuclear interactions with the RFQ copper structural material. This implies that there is no contribute to the dose rates outside the RFQ due to both secondary neutrons and prompt gammas. The maximum value of beam-off dose rate, due to the structural material activated by deuterons, is 5.6E-02 μSv/h at the surface of RFQ section 10 (5 MeV deuterons), value much less than 10 μSv/h, the maximum allowable dose rate limit for workers’ access.
In DTL, the last Tank 10 (highest deuteron energy, 40 MeV, and highest deuteron lost current, 246nA) gives the largest values of dose rates at beam-on and beam-off phases. The total beam-off dose rate, that is significant for the ORE, calculated considering a continuous irradiation scenario of 1 year, is 3.25 μSv/h at 1 m from the tank surface and at 1 week of cooling time. The area around the accelerator can be considered a restricted access area for maintenance, with specific procedures for worker protection
ITER Vacuum Vessel materials: impact of Cobalt content on the clearance index and contact dose rate
No full text0.05 wt% is the reference value of Cobalt content in the ITER Inboard and Outboard Vacuum Vessel materials. Different Co contents have been considered and the effects on the corresponding waste management policy and maintenance Occupational Radiation Exposure are presented and discussed.
The analyses hughlighted the impact of the cobalt content (from 0.01 up to 0.25 wt%) into the different Vacuum Vessel steels on the material clearance index of the Inboard and Outboard Vacuum Vessel zones (inner and outer shells and shielding plates) and on the material contact dose rate of the inner and outer shell of the Outboard Vacuum Vessel.
The ANITA-2000 activation code package, with the EAF-2003 neutron activation library, has been used for the calculations.
The same operating conditions used for the ITER GSSR reference calculation have been assumed.
The following conclusions have been outlined from the results analysis:
• the increasing of cobalt content on all the VV materials (both inboard and outboard VV zones) from 0.01 to 0.25 wt% does not change significantly their behavior from a waste management policy point of view, i.e. if a material containing 0.05 wt% Co (ITER reference) is clearable within 100 years from the final plant shut down it remains clearable within 100 years also with a 0.25 wt% Co content.
• the time needed to reach a material contact dose lower than 100 microSv/h at the Outboard Vacuum Vessel Front and Rear Walls depends clearly on the cobalt content in the 316L(N)-IG material of the corresponding zones. This is significant mainly for the Rear Wall: the time increases from about 4 months to about 8 years if Cobalt content increases from 0.01 to 0.25 wt%. For the Front Wall that time increases correspondingly from about 67 to about 82 years.
Detailed results of the activation calculation are given in appendix
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