Volume 78, Issue 5

Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s–early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative studies of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton® GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution.

Numerical simulation was performed to study the turbulent mixing behavior of two opposing flows inside a square chimney structure of a research reactor. The chimney design facilitates drawing pool water in the downward direction and thereby suppresses the upward flow of radioactive water jet to limit the radiation field at the reactor pool top. Analyses were carried out considering a mass flow rate of 750 kg/s for the upward flowing hot water from the core, which corresponds to Reynolds number of 3 × 10 6 . Mass flow ratios of the downward flow and the upward flow were 0.0, 0.05, 0.1 and 0.15. The effects of mass flow ratio, chimney height on the velocity and temperature distribution inside three-dimensional chimney structure was evaluated using CFD code PHOENICS. The effect of temperature difference between the opposing flows on velocity was also analysed. It is observed that increase in downward flow causes the jet height to decrease due to the opposing momentum of downward flow against the upward jet. The effects of chimney height and temperature difference on the jet height are found to be marginal because of dominating inertial force over buoyancy force for the study.

High Level Liquid Radioactive Waste (HLLRW) produced during reprocessing of spent fuel from nuclear reactors is encased in the canisters after vitrification. The vitrified waste has high heat generation rate due to decay heat and needs interim storage under surveillance. The waste needs to be cooled continuously until major portion of the decay heat is dissipated. Natural circulation air cooling has been considered to cool the canisters. Canisters are placed in a storage vault and cooled by induced axial flow of air with the help of stack. The capacity of storage vault for Vitrified Waste Storage Block (VWSB) Facility proposed at Integrated Plant-1, Tarapur is designed for interim storage of waste generated of 30 yrs of IP1 plant operation. Canister and concrete temperature should be within the prescribed limit. Parametric studies have been carried out for the relevant parameters such as stack and duct dimensions, plenum height etc. Details canister temperature have been obtained using CFD code CFD-ACE+. Axial and radial temperature variation in the canisters, thimble and ventilation pipe have been evaluated in a location. Effect of natural convection (in air) within the canister and between thimble and canister is also studied. It was found that canister centerline temperature reduces by 20°C.

The knowledge and understanding of water flow and solute transport in the unsaturated zone is becoming increasingly important especially in mitigation of groundwater pollution. Fate of radionuclide in the geological environment is a topic to address in performance and safety assessment studies for nuclear waste disposal and may be modeled considering flow and transport in porous media. However, often, due to the heterogeneity and anisotropy of the real systems, the computer simulations may be difficult to render the real behavior. This paper addresses the simulation of a tracer transport in the unsaturated zone of the Saligny site, the potential location for the Romanian low and intermediate level waste (LILW) disposal. Computation was based on experimental data and was performed with the Alliances platform, a numerical tool developed by French organizations CEA, ANDRA and EDF. In order to obtain information regarding the solute migration in depth and the solute lateral dispersion, the dispersivity coefficients of iodine were investigated in order to match the experimental concentration determined on samples from different locations of the site. A close fit of the simulation over experimental data for the water saturation profile at a depth of 0.5 m in transient state was targeted by taking into account evapotranspiration in order to obtain a realistic estimation of the water infiltration in the porous media. Dispersivity coefficients obtained from the simulation of the tracer transport are in good order of magnitude for the unsaturated area and allow to have a good preview of the tracer plume. However, further investigations are recommended on new samples in order to validate the migration of the tracer plume as expected.

Seasonally gross alpha–beta activities were measured in surface water and sediments in Sır Dam Pond (Kahramanmaras, Turkey). Six sampling sites were pre-defined in different locations of the Sır Dam Pond. Preliminary studies of gross-α and gross-β activities in the surface water and deep sediments were determined. The results obtained from the determination of the activity indicate that gross-α and gross-β activities were ranging from 0,007 ± 0,005 to 0,077 ± 0,006 Bq/l and from 0,028 ± 0,006 to 0,189 ± 0,009 Bq/l for the surface water and from 0,001 ± 0.001 to 0,229 ± 0,032 Bq/g and from 0,144 ± 0,026 to 0,419 ± 0,033 Bq/g for deep sediment, respectively.

Radioactive emissions of 13 lignite-fired power plants in Turkey are of great concern to the public and to scientists alike. The purpose of this study is to evaluate these power plants, according to their radioactive emissions by using the Analytic Hierarchy Process. Control criteria are in particular 226 Ra, 232 Th, 40 K and 238 U emissions from the power plants. These control criteria are weighted according to the objective assessment. The calculations are repeated for three different objective assessments of control criteria namely the mortality risk coefficients for inhalation, ingestion, external exposure of 226 Ra, 232 Th, 40 K and 238 U. It has been calculated that the Çan lignite-fired power plant is ranking first while the Soma-B plant is ranking last according to the radioactive emissions of the power plants when the average of three different objective control criteria are used in the calculations.

In this paper, a specific type of the Loss of Coolant Accident (LOCA), the DECL (Double Ended Cold Leg) break, that means totally guillotine type of break in the cold leg pipe, has been modeled. The accident is simulated with the CONTAIN 2.0 code. In the event of a LOCA accident, coolant mass and energy are released to the containment through the break. This causes an increase of pressure and temperature in the containment. The modeling is performed in the VVER-1000 reactor containment. The analysis includes average pressure in the containment and temperature distribution in sample cells in the long-time. The results are compared with the existing reports on studies that used the ANGAR code. Results show that the CONTAIN 2.0 code is an adaptable tool for the analysis of nuclear events such as DECL accident.

The criticality problem for one-speed neutrons in a uniform homogeneous slab with reflecting boundary conditions is studied using Henyey-Greenstein (HG) and Anlı-Güngör (AG) phase functions. The critical half-thicknesses of the slab are performed with traditional spherical harmonics (PN) method for various values of the cross-section parameter, reflection coefficient and the scattering parameters of HG and AG phase functions. The numerical results obtained in case of using both HG and AG phase functions are tabulated in the tables and they are compared with each other.

The effect of gamma radiation on sorption of 85 Sr 2+ on trace manganese dioxide and on precipitation of 131 I – with Ag + ions has been studied. The transition of both radionuclides into the solid phase by sorption and precipitation was found by measurement of the self-diffusion coefficients of 85 Sr 2+ and of 131 I – . The gamma radiation had negative effect on sorption and precipitation and also affected the ionic state of 85 Sr 2+ and of 131 I – .