Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 °C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 °C). Additionally, the peak drywell temperature of 155.3 °C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 °C, which is below the pool temperature used for evaluating the net positive suction head of pumps of the RHR system and the Emergency Core Cooling Systems (96.7 °C). The peak containment pressure and temperature are well below the design value (386.1 kPaG and 171.1 °C). Containment integrity of Chinshan Plant can be maintained under the SPU condition.     

A numerical and experimental study of water ingression phenomena in melt pool coolability

During a postulated severe accident, the core can melt and the melt can fail the reactor vessel. Subsequently, the molten corium can be relocated in the containment cavity forming a melt pool. The melt pool can be flooded with water at the top for quenching it. However, the question that arises is to what extent the water can ingress in the corium melt pool to cool and quench it. To reveal that, a numerical study has been carried out using the computer code MELCOOL. The code considers the heat transfer behaviour in axial and radial directions from the molten pool to the overlaying water, crust generation and growth, thermal stresses built-in the crust, disintegration of crust into debris, natural convection heat transfer in debris and water ingression into the debris bed. To validate the computer code, experiments were conducted in a facility named as core melt coolability (COMECO). The facility consists of a test section (200 mm × 200 mm square cross-section) and with a height of 300 mm. About 14 L of melt comprising of 30% CaO + 70% B₂O₃ (by wt.) was poured into the test section. The melt was heated by four heaters from outside the test section to simulate the decay heat of corium. The melt was water flooded from the top, and the depth of water pool was kept constant at around 700 mm throughout the experiment. The transient temperature behaviour in the melt pool at different axial and radial locations was measured with 24 K-type thermocouples and the steam flow rate was measured using a vortex flow meter. The melt temperature measurements indicated that water could ingress only up to a certain depth into the melt pool. The MELCOOL predictions were compared with the test data for the temperature distribution inside the molten pool. The code was found to simulate the quenching behaviour and depth of water ingression quite well.     

Applicability of small-scale integral test data to the 4500 MWt ESBWR loss-of-coolant accidents

This paper discusses the scaling methodology used by GE Hitachi Nuclear Energy (GEH) to show that the data obtained from the small-scale integral test facilities, namely, GIST and GIRAFFE-SIT, are relevant to the postulated loss-of-coolant accident (LOCA) of the 4500 MWt ESBWR. The conservation of mass and energy equations for the steam-water mixture in the reactor pressure vessel (RPV) are transformed to the equations for the rates of pressure change and water mass or inventory change. These equations are non-dimensionalized based on the most dominant physical processes of the individual stages of a LOCA, namely, the late blowdown stage, the GDCS (gravity driven cooling system) transition stage and the full GDCS stage. The magnitudes of the non-dimensional Pi-groups, obtained from these equations, for the 4500 MWt ESBWR are compared with those obtained for the small-scale integral tests mentioned earlier. In addition, simplified analyses were conducted for the first two stages by integrating the non-dimensional RPV depressurization rate and the water inventory change rate equations. The results of the 4500 MWt ESBWR are very similar to the test data obtained from the GIST and the GIRAFFE-SIT test facilities. Therefore, based on both the Pi-group magnitudes and the simplified analyses, it is concluded that the small-scale integral test data mentioned above are applicable to the 4500 MWt ESBWR LOCA applications.     

Monte Carlo simulation of a conceptual thermal column in the Syrian miniature neutron source reactor using MCNP-4C

Based on probabilistic approach, the MCNP-4C code has been used effectively to simulate the Syrian MNSR reactor core and all its surrounding components in three dimensions, including a preliminary conceptual design of a thermal column to be installed later. For verification and validation purposes, reactor calculations include: criticality and control rod worth. Values of these parameters are 1.00517 and 6.54 mk, respectively. The thermal column is to be installed in the water of the reactor pool. Optimal conditions for this thermal column were tested using the already developed model. Optimization focused on the most suitable position for placement of the column in the water pool, dimensions, and material. The aim was to have a thermal neutron flux of 1 × 10⁹ n cm⁻² s⁻¹ in the center of thermal column, and resonant and fast neutron fluxes to be as low as possible as well.     

Lead-cooled flexible conversion ratio fast reactor

Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO₂ (S-CO₂) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO₂ PCS.     

Investigation of appropriate mesh size and solid angle number for CFD simulating the characteristics of pool fires_with experiments assessment

In order to implement NFPA 805 in the performance-based fire design for nuclear power plants (NPPs), a reliable computational fluid dynamics (CFD) fire model is needed. However, numerical treatments including mesh size and number of solid angles significantly influence the accuracy of CFD predicted results of the thermal-hydraulic behaviors involved in fire scenarios. Therefore, the majority of this paper is to investigate appropriate mesh size and solid angle number used for CFD simulating the characteristics of pool fires. Based on the present sensitivity studies of different mesh sizes and solid angle numbers, appropriate numerical treatment could be selected by comparing the predicted results of flame height and radiative heat flux for the pool fires. Two experiments with pan sizes of 20 and 30 cm, respectively, are also conducted to assess the CFD predicted results obtained using these selected mesh size and number of solid angles. Good agreement between the experiments and predictions clearly shows that the optima mesh resolution for the flame height and radiative heat flux is at the normalized mesh size (R*, ratio of the mesh size to the flame characteristic diameter) = 0.077 and 500 solid angles for thermal radiation are sufficient to reasonably capture the radiative characteristics from the pool fires with the size less than 30 cm. Herein, solid angle means included angle between rays from thermal radiator.     

Dynamic model of Fast Breeder Test Reactor

Fast Breeder Test Reactor (FBTR) is a 40 M Wt/13.2 MWe sodium cooled reactor operating since 1985. It is a loop type reactor. As part of the safety analysis the response of the plant to various transients is needed. In this connection a computer code named DYNAM was developed to model the reactor core, the intermediate heat exchanger, steam generator, piping, etc. This paper deals with the mathematical model of the various components of FBTR, the numerical techniques to solve the model, and comparison of the predictions of the code with plant measurements. Also presented is the benign response of the plant to a station blackout condition, which brings out the role of the various reactivity feedback mechanisms combined with a gradual coast down of reactor sodium flow.     

PASCAR: Long burning small modular reactor based on natural circulation

PASCAR is a 100 MWt/35 MWe lead-bismuth-cooled small modular reactor which requires no on-site refueling and well suits to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. This paper includes both steady-state and transient performance evaluations for neutronics and thermal-hydraulics. Through design optimization studies for minimizing a burn-up reactivity loss, the metallic fuels-loaded core was designed with less than 1$ reactivity swing over 20-year cycle. A radial peaking power location shows the slow inward migration from outer enrichment zones while maintaining peaking factor within 1.35, reducing radiation damage and corrosion duty of high temperature environments. Equipped with coolant flow path large enough to ensure low pressure drop, this reactor is intended to operate by only natural circulation of chemically inert coolant within relatively low temperature range, 320-420 °C. Peak outlet temperature is nearly 450 °C where an Al-containing duplex cladding has sufficient corrosion resistance. Despite of 50% decrease of fuel thermal conductivity after swelling, inherent negative reactivity feedback and passive decay heat removal capability could secure an ample safety margin of peak fuel centerline temperature in tow safety analyses, unprotected transient overpower and unprotected loss of heat sink. The likelihood of loss of coolant, loss of flow, and local blockage is virtually eliminated by employing respectively a double-walled vessel, pump-less cooling, and cross-flow allowed open square assemblies. Simple fabrication, modular construction, and long burning cycle would compensate for economic disadvantages over smaller power and lower temperature than those of conventional fast reactors.     

Experimental study on the air/water counter-current flow limitation in a model of the hot leg of a pressurized water reactor

An experimental investigation on the air/water counter-current two-phase flow in a horizontal rectangular channel connected to an inclined riser has been conducted. This test-section representing a model of the hot leg of a pressurized water reactor is mounted between two separators in a pressurized experimental vessel. The cross-section and length of the horizontal part of the test-section are (0.25 m × 0.05 m) and 2.59 m, respectively, whereas the inclination angle of the riser is 50°. The flow was captured by a high-speed camera in the bended region of the hot leg, delivering a detailed view of the stratified interface as well as of dispersed structures like bubbles and droplets. Countercurrent flow limitation (CCFL), or the onset of flooding, was found by analyzing the water levels measured in the separators. The counter-current flow limitation is defined as the maximum air mass flow rate at which the discharged water mass flow rate is equal to the inlet water mass flow rate.From the high-speed observations it was found that the initiation of flooding coincides with the formation of slug flow. Furthermore, a hysteresis was noticed between flooding and deflooding. The CCFL data was compared with similar experiments and empirical correlations available in the literature. Therefore, the Wallis-parameter was calculated for the rectangular cross-sections by using the channel height as length, instead of the diameter. The agreement of the CCFL curve is good, but the zero liquid penetration was found at lower values of the Wallis parameter than in most of the previous work. This deviation can be attributed to the special rectangular geometry of the hot leg model of FZD, since the other investigations were done for pipes.     

Influence of elbow curvature on flow structure at elbow outlet under high Reynolds number condition

In the design of Japan Sodium-cooled Fast Reactor (JSFR), coolant velocity is beyond 9 m/s in the primary hot leg pipe of 1.27 m diameter. The Reynolds number in the piping reaches 4.2 × 10⁷. Moreover, a short-elbow is adopted in the hot leg pipe in order to achieve compact plant layout and to reduce plant construction cost. Therefore, the flow-induced vibration (FIV) arising from the piping geometry may occur in the short-elbow pipe. The FIV is due to the excitation source which is caused by the pressure fluctuation in the pipe. The pressure fluctuation in the pipe is closely related with the velocity fluctuation. As the first step of clarification of the FIV mechanism, it is important to grasp the mechanism of flow fluctuation in the elbow. In this study, water experiments with two types of elbows with different curvature ratios were conducted in order to investigate the interaction between flow separation and the secondary flow due to the elbow curvature. The experiments were conducted with the short-elbow and the long-elbow under Re = 1.8 × 10⁵ and 5.4 × 10⁵ conditions. The velocity fields in the elbows were measured using a high-speed Particle Image Velocimetry (PIV). The time-series of axial velocity fields and the cross-section velocity fields obtained by the high-speed PIV measurements revealed the unsteady and complex flow structure in the elbow. The flow separation always occurred in the short-elbow while the flow separation occurred intermittently in the long-elbow case. The circumferential secondary flows in clockwise and counterclockwise directions flowed forward downstream of reattachment point alternately in both elbows.     

An innovative research reactor design

A new and innovative core design for a research reactor is presented. It is shown that while using the standard, low enriched uranium as fuel, the maximum thermal flux per MW of power for the core design suggested and analyzed here is greater than those found in existing state of the art facilities without detrimentally affecting the other design specs. A design optimization is also carried out to achieve the following characteristics of a pool type research reactor of 10 MW power: high thermal neutron fluxes; sufficient space to locate facilities in the reflector; and an acceptable life cycle. In addition, the design is limited to standard fuel material of low enriched uranium. More specifically, the goal is to maximize the maximum thermal flux to power ratio in a moderate power reactor design maintaining, or even enhancing, other design aspects that are desired in a modern state of the art multi-purpose facility. The multi-purpose reactor design should allow most of the applications generally carried out in existing multi-purpose research reactors. Starting from the design of the German research reactor, FRM-II, which delivers high thermal neutron fluxes, an azimuthally asymmetric cylindrical core design with an inner and outer reflector, is developed. More specifically, one half of the annular core (0 < θ < π) is thicker than the other half. Two variations of the design are analyzed using MCNP, ORIGEN2 and MONTEBURNS codes. Both lead to a high thermal flux zone, a moderate thermal flux zone, and a low thermal flux zone in the outer reflector. Moreover, it is shown that the inner reflector is suitable for fast flux irradiation positions. The first design leads to a life cycle of 41 days and high, moderate and low (non-perturbed) thermal neutron fluxes of 4.2 × 10¹⁴ n cm⁻² s⁻¹, 3.0 × 10¹⁴ n cm⁻² s⁻¹, and 2.0 × 10¹⁴ n cm⁻² s⁻¹, respectively. Heat deposition in the cladding, coolant and fuel material is also calculated to determine coolant flow rate, coolant outlet temperature and maximum fuel temperature under steady-state operating conditions. Finally, a more compact version of the asymmetric core is developed where a maximum (non-perturbed) thermal flux of 5.0 × 10¹⁴ n cm⁻² s⁻¹ is achieved. The core life of this more compact version is estimated to be about 23 days.     

Experimental investigation of weir instability in main vessel cooling system of 1/4 FBR model

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam, India. The main vessel of this pool type reactor acts as the primary containment in the reactor assembly. In order to keep the main vessel temperature below creep range and to reduce high temperature embrittlement and also to ensure its healthiness for 40 years of reactor life, a small fraction of core flow (0.5 m³/s) is sent through an annular space formed between the main vessel and a cylindrical baffle (primary thermal baffle) to cool the vessel. The sodium after cooling the main vessel overflows the primary baffle (weir shell) and falls into another concentric pool of sodium separated from the cold pool by the secondary thermal baffle and then returned to cold pool. The weir shell, where the overflow of liquid sodium takes place, is a thin shell prone to flow induced vibrations due to instability caused by sloshing and fluid-structure interaction. A similar vibration phenomenon was first observed during the commissioning of Super-Phenix reactor. In order to understand the phenomenon and provide necessary experimental back up to validate the analytical models, weir instability experiments were conducted in a 1:4 scale stainless steel (SS) model installed in a water loop. The experiments were conducted with flow rate and fall height as the varying parameters. The experimental results showed that the instability of the weir shell was caused due to fluid structure interaction. This paper discusses the details of the model, the modeling laws, similitude criteria adopted, analytical prediction, the experimental results and conclusion.     

Physics design of experimental metal fuelled fast reactor cores for full scale demonstration

Fast breeder reactors based on metal fuel are planned to be in operation for the year beyond 2025 to meet the growing energy demand in India. A road map is laid towards the development of technologies required for launching 1000 MWe commercial metal breeder reactors with closed fuel cycle. Construction of a test reactor with metallic fuel is also envisaged to provide full-scale testing of fuel sub-assemblies planned for a commercial power reactor. Physics design studies have been carried out to arrive at a core configuration for this experimental facility. The aim of this study is to find out minimum power of the core to meet the requirements of safety as well as full-scale demonstration. In addition, fuel sustainability is also a consideration in the design. Two types of metallic fuel pins, viz. a sodium bonded ternary (U-Pu-6% Zr) alloy and a mechanically bonded binary (U-Pu) alloy with 125 μm thickness zirconium liner, are considered for this study. Using the European fast reactor neutronics code system, ERANOS 2.1, four metallic fast reactor cores are optimized and estimated their important steady state parameters. The ABBN-93 system is also used for estimating the important safety parameters. Minimum achievable power from the converter metallic core is 220 MWt. A 320 MWt self-sustaining breeder metal core is recommended for the test facility.     

Crystal structure, corrosion kinetics of new zirconium alloys and residual stress analysis of oxide films

Corrosion kinetics of NZ2 alloy were investigated after autoclave treatments in 360 °C/18.6 MPa lithiated water and 400 °C/10.3 MPa steam. The crystal structure and the residual stress of oxide films of NZ2 alloys after corroded in both conditions were investigated by XRD method. The kinetics analysis indicates that the resistance of NZ2 alloy treated in 360 °C lithiated water is higher than that treated in 400 °C steam. The crystal structure analysis shows that the content of tetragonal t-ZrO₂ in the oxide films decreases smoothly and the content of monoclinic m-ZrO₂ increases with the duration of corrosion time, independent of the kinetics transition. Stress measurements show that high compressive stresses were developed in the oxide layers. Furthermore, the transitions of kinetics can be associated with the sudden decrease of macroscopic compressive stresses in the oxide films. The higher t-ZrO₂ content is, the higher compressive stress in the oxide film is, the lower is the corrosion rate. Therefore it is considered that t-ZrO₂ is mainly stabilized by the macroscopic compressive stresses in the oxide films. In addition, local stresses in the oxide films, grain size and the oxygen vacancies play an important role in the t-ZrO₂ stabilization.     

Erratum to “Experimental study on the air/water counter-current flow limitation in a model of the hot leg of a pressurized water reactor”

This paper replaces the paper published in the journal by Deendarlianto et al. (2008). Because of an error in the implementation of the air flow meter some of the data given by Deendarlianto et al. (2008) are wrong. They are corrected within the present paper. The general results and conclusions remain unchanged.An experimental investigation on the air/water counter-current two-phase flow in a horizontal rectangular channel connected to an inclined riser has been conducted. This test-section representing a model of the hot leg of a pressurized water reactor is mounted between two separators in a pressurized experimental vessel. The cross-section and length of the horizontal part of the test-section are (0.25 m × 0.05 m) and 2.59 m, respectively, whereas the inclination angle of the riser is 50°. The flow was captured by a high speed camera in the bended region of the hot leg, delivering a detailed view of the stratified interface as well as of dispersed structures like bubbles and droplets. Counter-current flow limitation (CCFL), or the onset of flooding, was found by analyzing the water levels measured in the separators. The counter-current flow limitation is defined as the maximum air mass flow rate at which the discharged water mass flow rate is equal to the inlet water mass flow rate.From the high-speed observations it was found that the initiation of flooding coincides with the formation of slug flow. Furthermore, a slight hysteresis was noticed between flooding and deflooding. The CCFL data was compared with similar experiments and empirical correlations available in the literature. Therefore, the Wallis-parameter was calculated for the rectangular cross-sections by using the channel height as length, instead of the diameter. The agreement of the CCFL curve is good, but the zero liquid penetration was found at lower values of the Wallis parameter than in most of the previous work. This deviation can be attributed to the special rectangular geometry of the hot leg model of HZDR, since the other investigations were done for pipes.     

Parametric analyses of partial sub-assembly blockages of the 50 MWth gas cooled experimental technology and demonstrator reactor (ETDR)

This paper presents a parametric study of thermal hydraulic and structural mechanic analyses of accidental blockages of the hottest sub-assembly of the 50 MWth gas-cooled fast reactor, ETDR. The blockage ratios were 60% and 100% of the sub-assembly flow cross-section located at the first row of grid spacers. Temperature profiles in the fuel and cladding were calculated as a function of time using computational fluid dynamics. The results were incorporated into finite elements analyses to evaluate thermal and mechanical stresses and strains in the cladding and fuel. 2D simulations with generalized plane strains were used in the structural analyses applying the maximum power density in the pellet as calculated by CFD. The thermal analyses showed that a 60% partial blockage increases the maximum cladding temperature by 130 °C within a time period of 50 s, whereas a full blockage will lead to clad melting (1320 °C) in about 8 s after accident initiation. In the finite element analyses several, mainly conservative, assumptions have been made to incorporate the phenomena occurring in the pellet-cladding interaction. The results of the finite element analyses represent a first study of the pellet-cladding mechanical interface under given transients, exploring the development of a methodology to be used for future analyses. The model has been verified to predict realistic contact stresses in line with analytical solutions during partial or full blockages. However, more elaborate 2D and 3D contact models, with creep and irradiation creep material models for both fuel and clad, together with parametric studies on friction coefficients and number of fuel fragments are foreseen for future work on failure criteria.     

Deuterium retention of boron-titanium and reduction of deuterium retention after helium ion irradiations

Deuterium ion irradiations with an ion with energy of 1.7 keV were conducted for boron-titanium (B-Ti) film prepared by electron beam evaporation and hot pressed titanium-boride, TiB₂. The amount of retained deuterium was measured for these materials using a technique of thermal desorption spectroscopy. The amount of deuterium retained in TiB₂ was comparable with that in B-Ti. Desorption peaks of deuterium in B-Ti were 470 K and 620 K, corresponding to a desorption in the low temperature regime observed in boron (B) and a desorption in titanium (Ti), respectively. The desorption peaks in TiB₂ were 620 K and 750 K, which correspond to the desorption in Ti and that in the high temperature regime in B, respectively. The desorption temperature in B-Ti was approximately 100 K lower than that in TiB₂. This difference is discussed based upon chemical bindings and amorphous/crystal structures of B-Ti and TiB₂. Irradiation of helium ion with energy of 5 keV was conducted for B-Ti after the deuterium ion irradiation. The amount of retained deuterium decreased and the desorption temperature shifted to the lower temperature regime, as the helium ion fluence increased. The shift to the low temperature regime is due to the enhancement of amorphous structure of B in B-Ti.     

Physics design of initial and approach to equilibrium cores of a reactor concept for thorium utilization

A thermal reactor concept ‘a thorium breeder reactor’ (ATBR) was conceived and reported by the authors during 1998. The distinctive physical characteristics of ATBR core with different types of seed fuels have been discussed in subsequent publications. The equilibrium core of ATBR with Pu seed was shown to exhibit a flat and low excess reactivity for a fuel cycle duration of two years. Notably this is achieved by no conventional burnable poison but by intrinsic balancing of reactivity between fissile and fertile zones. In this paper we present the design of the initial core and the refueling strategy for subsequent fuel cycles to enable a smooth transition to the equilibrium core. Three fuel types with characteristics similar to the three batch fuels of equilibrium core were designed for the initial core. Fuel requirement for the initial core is 4673 kg of reactor grade (RG) Pu for a cycle length of two years at 1875 MWt as against the 2200 kg needed for each fuel cycle of equilibrium core for same quantum of energy. The core reactivity variation during the first fuel cycle is monotonic fall and is relatively high (∼40 mk) but gradually diminishes to ±5 mk for fuel cycles 5–8.     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

Rapid and direct micro-machining/patterning of polymer materials by oxygen MeV ion beam irradiation through masks

PTFE (PolyTetraFluoroEthylene), often called Teflon, is a well-known polymer for being a non-stick material with good thermal properties. Moreover, PTFE is biocompatible and especially it is a cyto-compatible polymer. To enable bonding, a chemical etching based on sodium solutions is generally used to modify surfaces. In this paper we study the etching of PTFE using an oxygen ion beam in the MeV energy range. We present micro-patterning of PTFE through masks with two fluences of 5 × 10¹⁵ and 1 × 10¹⁶ ion cm⁻². As is demonstrated the use of a mask allows structuring of large areas while maintaining a distance between the mask and sample makes industrial applications possible.