05 Nuclear fuels (scientific, technical)

the neutron treatment. The full recovering of the structure after isochronal annealing in vacuum is observed in all implanted specimens at the temperature of 450°C. Using the PLEPS technique, for the first time, depth profiling of the positron lifetime spectra in the near-surface region (20-500 nm) of hydrogen implanted copper alloys was performed and compared with the TRIM calculations and transmission electron microscopy (TEM) studies. Possible annihilation channels in CuCrZr and CuA125 materials are discussed in details together with corresponding annihilation characteristics determined theoretically and using computer simulations. The paper discusses the results of positron lifetime measurements of the irradiated and non-irradiated CuZrCr and CuA125 specimens, and ideniffies the most probable types of positron trapping sites. Finally, the results are discussed in terms of microstructural changes of the studied materials upon irradiation and subsequent heat treatment.

05100105 Power stabilization and temporal performance of a peaceful nuclear explosion reactor with a mixture of 90% flibe +10% UF4 (or ThF4) Unalan, S. Fusion Engineering and Design, 2004, 70, (3), 233-246. This work investigated the power stabilization and the temporal neutronic behaviour of a peaceful nuclear explosion reactor (PACER) with ThF4 and UF 4 which produces an electrical energy of 1.2 GW from fusion explosions of 8.13x1012 J to be repeated every 40 min during the operation period of 30 year. The use of ThF4 and UF4 is realized by a mixture zone consisted of flibe and fuel, instead of full flibe zone. The mixture compositions determined by volume fraction are 90% flibe +10% UF4, 90% flibe +10% ThF4 and 90% flibe +5% UF4+5% ThF4. The capacity factor of the reactor is 0.75. The cylindrical explosion chamber has a radius of 30 m and a height of 75 m. The mixture mass of 18 000 tonnes having a zone thickness of 5 m were circulated during the operation period. The mixture zone would be subdivided into jets so that the gas and the vapour bypasses the liquid as it vents and does not accelerate the liquid mixture to high velocities. The selected volume fraction is 75% void +25% mixture. The use of fuel materials in the PACER reactor resulted in high-energy production, sufficient tritium breeding and significant fissile fuel breeding. The averages of tritium breeding ratio (TBR) values over 30 years are between 1.1 and 1.17. Generally, the mixtures with UF4 show better performance than the mixture with ThF4. For the mixtures with ThF4, ThF4+UF4 and UF4, the energy production without the separation process reached from ~14311 MW (electric), ~1700 MW (electric) and ~2000 MW (electric) to ~1900 MW (electric), ~2150 MW (electric) and ~2320 MW (electric), respectively. The reached cumulative fissile fuel enrichments in the fuel (CFFE) in percentage are 1.8, 2.45 and 2.4%, respectively. The fuel obtained from the PACER could be used as a nuclear fuel only in the CANDU and the advanced CANDU. In addition, the stabilization process is performed by means of the plutonium or uranium fuel separation from the mixture, after the energy output of the reactor reaches 1600 MW (electric), 1800 MW (electric) and 2000 MW (electric) at the operation periods of 11, 6 years and 2 months, respectively. At the end of the separation process, the separated fuel amounts are about 15,374 and 11 tonnes, respectively. Tile CFFE values of the separated fuel at the end and at the start up of the separation process are 99.36 and 99.23%, 1.13 and 3.9%, and 99.99 and 99.2%, respectively. The CFFE values of tile remained fuel at the end of the separation process are ~0.7, and 2.2%, ~0.7%, respectively. Consequently, in the evaluation in terms of sufficient tritium breeding, high energy production, significant fissile fuel production and the nuclear weapon hazard of the fuel, the mixture of 90% flibe +5% ThF4+5% UF4 exhibited the highest performance.

05/00106 Remote experiment participation on Tore-Supra Theis, J.-M. and Larsen, J.-M. Fusion Engineering and Design, 2004, 71, (1-4), 257 261. The DRFC has traditionally had a very large external collaboration involvement. In particular, 15% of the DRFC work is directed towards the JET programme. As a consequence substantial telecommunications facilities have been installed. A specific station for remote communi- cation has been set up in the Tore-Supra control room, closely coupled to a collaborating team at INRS Que., Canada. This paper describes the pilot experience with the Canadian participation, which gives details of the communication and data sharing tools used to fully work on Tore-Supra.

05/00107 Shield structure optimisation studies for the west beam port of the KAMINI reactor Sunny, C. S. and Subbaiah, K. V. Annals of Nuclear Energy, 2004, 31, (12), 1403-1413. KAMINI is the Kalpakkam Mini Reactor and its main purpose is to cater to the experimental needs and for neutron radiography. It is a water-cooled reactor with 233U as the fissile material. The reactor has three neutron beam ports for experimental needs, which are made of

16 Fuel and Energy Abstracts January 2005

graded cylindrical aluminium channel. All the beam ports start from the core, pierce the biological shield and are 2 m long. The shield structure optimization studies for the beam port towards the west side of the reactor are presented here. The diameter of the west beam channel at the core centre is 54 mm and at the other end is 250 mm. The west beam tube opening is 530 mm below the floor level and hence the pit housing the experimental cavity is below the floor level with dimensions 2 m x 2.5 m x 1.3 m. The beam tube opening into the experimental cavity serves as the neutron source for radiation physics experiments and is assumed as a surface source in the calculations. Rough estimate of the shield design is made based on the literature on dose-equivalent index transmission through concrete for average neutron energy of 1.5 MeV. Detailed radiation transport calculations are performed using Monte Carlo neutral particle transport code (MCNP) to optimize the shield design. Neutron and capture gamma dose rates at the accessible areas are estimated. The contribution of prompt fission gamma rays is found to be negligible compared to the dose rates due to capture gamma rays. The details of the optimized shield structure proposed for the west beam port are as follows. Fixed concrete shields of thickness 650 mm on the lateral sides and a composite shield (500 mm paraffin and 50 mm concrete) in the front side at a distance of 1 m from the beam tube opening are recommended inside the experimental pit. During reactor operation, a composite mobile shield (500 mm paraffin and 500 mm concrete) closes the experimental cavity at the floor level. Fixed concrete shields are recommended to close the pit fully. The shield structure3proposed increases the experimental cavity volume from 0.2 to 1.4 m with the dose levels at the accessible areas less than one laSv/h. The MCNP computed neutron and gamma dose rates are compared with the measured values with the existing shield structure to verify the source term used.

05/00108 Studies on the corrosion behavior of cerium- implanted zirconium Peng, D. Q. et al. Journal ~?f Nuclear Mater/a/s, 2004, 324, (l), 71-75. In order to study the influence of cerium ion implantation on the aqueous corrosion behaviour of zirconium specimens were implanted with cerium ions with a fluence ranging from 1 × 102 to 1 x l0 "21 ions/ m 2 at about 150°C, using a MEVVA source at an extracted voltage of 40 kV. The valence and element penetration distribution of the surface layer were analysed by X-ray photoelectron spectroscopy (XPS) and auger electron spectroscopy (AES) respectively. The potentiodynamic polarization technique was employed to investigate the aqueous corrosion resistance of zirconium in a 1 N H2SO4 solution. It was found that there was a remarkable improvement in the aqueous corrosion behaviour of zirconium implanted with cerium ions com- pared with that of the as-received zirconium. The corrosion resistance improvement of the cerium-implanted zirconium is probably due to the addition of cerium oxide dispersoid into the zirconium matrix and oxidization protection.

05•00109 Thermo-hydrodynamic design and safety parameter studies of the TRIGA MARK II research reactor Huda, M. Q. and Rahman, M. Annals c~fNuclear Energy, 2004, 31, (10), 1101-1118. The PARET computer code was used to analyse important thermo- hydrodynamic design and safety parameters of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The study involves the determi- nation of the departure from nucleate boiling (DNB) value and studying its effect over the thermo-hydrodynamic design of the reactor. In the process the temperature profile, heat flux and pressure drop across the hottest channel of the TRIGA core were evaluated. The DNB ratio (DNBR), which is defined as the ratio of the critical heat flux to the heat flux achieved in the core, was computed by means of a suitable correlation as defined in PARET code. Over the length 0.381 m of the hottest channel the DNBR varies, starting from 3,8951 to 5.4031, with a minimum of 2.7851. The peak heat flux occurs at the axial centre of the fuel elements; therefore the DNBR is minimum at this location. The reactor core should be designed so as to prevent the DNBR from dropping below a chosen value under a high heat flux transient condition for the most adverse set of mechanical and coolant conditions. The loss-of-flow accident (LOFA) scenario of the reactor has also been studied to ensure that the existing design and procedures are adequate to assure that the consequences from this anticipated occurrence does not lead to a significant accident. The loss-of-flow transient after a trip time of 4.08 s at 85% of loss of normal flow for the TRIGA core shows a peak temperature of 709.22°C in the fuel centreline and 131.94°C in the clad and 46.63°C in the coolant exit of the hottest channel. The transient was terminated at 15% of nominal flow after approximately 48.0 s and the time at which the reversal of coolant flow starts is approximately 67.0 s.