INR Pitesti Research Articles

Refining the Welding Technique for Nuclear Fuel Elements at INR Pitesti

Sealing of nuclear fuel material inside the fuel element clad must create a leak-tight "safety barrier" to prevent the release of radioactive products to environment. High quality welds are mandatory to withstand harsh conditions (radiation, pressure, temperature, corrosion) making possible the safe operation of nuclear reactors. The joint design, material selection and welding technique must be combined by smartly balancing possible technological options to yield the best attainable quality for the intended purpose; these choices are discussed in the paper. For thin-walled clad to end-plug welding, heat flow pattern as determined by joint design and fitting accuracy proved to be crucial for the fusion boundary shape and moreover for the success rate in automate welding. Consequently, Finite Element Analysis of the transient thermal field during welding was performed, in order to determine the best compromise with reasonable machining precision for parts. The main features of the developed thermal model and some results illustrating its good predictions vs. actual welds are also presented. Helium-shielded pulsed welding was initially preferred to minimize HAZ, distortion and porosity but unfortunately important cast-to-cast variation in penetration was observed with Inconel-600 plugs, due to Marangoni effect. Extensive work was done to overcome this, mainly through variation of pulsing and of the shielding gas; depth-to-width ratio can be noticeably improved with no material addition. Out of welding classic cladding materials, studies were initiated at INR on joining oxide-dispersion strengthened (ODS) alloys and specialty austenitic formulations (e.g. 15/15Ti) since they are candidate materials of great interest for the next generation of nuclear reactors.

Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

Abstract Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO2 grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation.

Investigation of the Ru-43LV fuel behaviour under LOCA conditions in a CANDU reactor

Abstract Presently, INR Pitesti is developing an advanced fuel design RU-43LV (recovered uranium fuel bundle with 43 elements and low void reactivity feature) based on recovered uranium from LWR. Compared with the current design of 37 natural uranium element (NU-37) fuel bundle, RU-43LV will have higher power capability and higher burn-up potential in CANDU reactors of Cernavodâ-Romania Nuclear Power Plant (NPP). Fuel burn-up of RU-43LV fuel will be about two times the burn-up usually achieved in CANDU reactors fuelled with natural uranium fuel. The effect of the design changes of RU-43LV bundle on the reactor safety has been analyzed and the results are presented in this paper. As part of the conceptual design study, the performance of the RU-43LV fuelled core during a large loss-of-coolant accident (LLOCA) was assessed with the use of several computer codes. The most relevant calculations performed regarding RU-43 RV fuel safety are presented in this paper. Also, the stages of an experimental program aiming to study RU-43LV fuel behaviour in high temperature transients are briefly described.

CANDU fuel elements behaviour in the load following tests

Abstract Two load following (LF) tests on CANDU type fuel elements were performed in TRIGA Research Reactor of INR Pitesti. In the first LF test the 78R fuel element has successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, the fuel element withstood 200 power cycles from 27 to 54 kW/m average linear power as well as additional ramps due to reactor trips and restarts during test period. New LF tests are planed to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in TRIGA Research Reactor and their relation to CANDU fuel performance in LF conditions.

In-pile measurements of the fission gas pressure and comparison with ROFEM code results

Abstract Two experimental fuel elements with different UO2 grain size were irradiated in the C2 capsule in the TRIGA MT reactor of INR Pitesti to a burn-up of 178.9 MWh kg−1 U and linear powers between 45 to 60 kW m−1. Both elements were instrumented to measure internal gas pressure during irradiation. Experimental results presented in this paper represent data relative to two fuel elements with different UO2 grain size. The predictions of fuel performance code ROFEM in terms of internal gas pressure were compared with the experimental data. Experimental programs are in progress at INR Pitesti. New instrumented tests are planed to be performed in the C2 capsule in order to establish the pellet microstructure influence on fission gas release. This paper describes measuring techniques used at INR Pitesti and present experimental results of the in-pile tests comparatively with the ROFEM code predictions.

In-reactor measurements of fuel centerline temperature variation during power change

Abstract The thermal response of the experimental fuel elements is examined by thermocouples placed at the centre of the fuel. Such measurements allow the determination of thermal feedback effects induced by simultaneous liberation of fission gases and pellet thermal expansion densification, swelling and relocation of fuel fragments. This paper describes briefly measuring techniques developed and currently used in INR Pitesti, and presents and discusses selected experimental results obtained during in-pile experiments performed in the TRIGA SS reactor of INR Pitesti in order to measure the fuel centerline temperature variation during power change.

Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

Abstract Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g–1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti.

The SANS facility at the Pitesti 14MW TRIGA reactor

The SANS facility existing at the Pitesti 14MW TRIGA reactor is presented. The main characteristics and the preliminary evaluation of the installation performances are given. A monochromatic neutron beam with 1.5 A ≤ λ ≤ 5 A is produced by a mechanical velocity selector with helical slots. A fruitful partnership was established between INR Pitesti (Romania) and JINR Dubna (Russia). The first step in this cooperation consists in the manufacturing in Dubna of a battery of gas-filled positional detectors devoted to the SANS instrument.

Scientific, technical

The CANDU type reactors have the ability to accommodate a wide variety of fuel types. For all the used fuel types the void coefficient of the reactivity is positive. In our paper a heterogeneous two-stratified coolant model is used for SEU-43 fuel bundle type. This is slightly enriched uranium (near 1%) with 43 fuel rods developed in INR Pitesti. The coolant is treated as a two-phase (liquid and vapour) medium, gravitationally separated. Eigenvalues are computed with the transport code CP_2D at different void fractions and burnup points. For the fresh fuel the MCNP results on the two-startified model a homogeneous model is used to benchmark CP_2D results.

Void reactivity and pin power calculation for a CANDU cell using the SEU-43 fuel bundle

The CANDU type reactors have the ability to accommodate a wide variety of fuel types. For all the used fuel types the void coefficient of the reactivity is positive. In our paper a heterogeneous two-stratified coolant model is used for SEU-43 fuel bundle type. This is slightly enriched uranium (near 1%) with 43 fuel rods developed in INR Pitesti. The coolant is treated as a two-phase (liquid and vapour) medium, gravitationally separated. Eigenvalues are computed with the transport code CP_2D at different void fractions and burnup points. For the fresh fuel the MCNP results on the two-startified model a homogeneous model is used to benchmark CP_2D results.

Spent Fuel Transport in Romania by Road: an Approach Considering Safety, Risk and Radiological Consequences

AbstractThe transport of high-level radioactive wastes, involving Type B package, is a part of the safety of the Romanian waste management programme and the overall aim of this activity is to promote the safe transport of radioactive materials in Romania. The paper presents a safety case analysis of the transport of a single spent fuel CANDU bundle, using a Romanian built Type B package, from the CANDU type nuclear power plant (NPP) Cernavoda to the INR Pitesti, in order to be examined within INR's hot-celss facilities. The safety assessment includes the following main aspects: (1) evaluaiion and analysis of available data on road traffic accidents; (2) estimaiion of the expected frequency for severe road accident scenarios resulting in potential radionuclide release; and (3) evaluation of the expected radiological consequences and accident risks of transport operations.

Qualification Tests of Packages used for Transport and Storage of Low Activity Radioactive Wastes In INR Pitesti

AbstractRadioactive wastes generated by the TRIGA INR research reactor are packaged according to the national and international rules and standards. The technology for packaging and treatment of radioactive wastes can also be used at the Nuclear Power Plant Cernavoda. The qualification tests for the package used for transport and storage of radioactive wastes (low activity, up to 6·07 GBq (0·164 Ci) per drum) are described. The package used is a drum manufactured from 1 mm thick mild steel with the dimensions: height 915 ± 10 mm; diameter 600 ± 5 mm; volume 220 litres. To achieve adequate safety in the transport of radioactive wastes strict precautions must be taken to protect transport workers, the general public and the environment by ensuring, both in normal and in accident conditions, adequate containment and shielding of the materials, according to the IAEA Regulations for the Safe Transport of Radioactive Materials. The adequacy of the package design is therefore of primary importance, the design re...

Qualification Tests for Packages used for Transport and Storage of Radioactive Waste (Low Activity) in INR Pitesti

AbstractRadioactive wastes generated by the TRIGA INR research reactor are packaged according to the national and international standards and the IAEA Regulations. The technology for packaging and treatment of radioactive wastes used in this institute can be applied, prospectively, at the Nuclear Power Plant Cernavoda, after commissioning. The qualification tests (type tests) are described for packages used for transport and storage (for a long period of about 30 years) of radioactive wastes (low activity, up to 0.5068×1010 Bq per drum, or 0.164 Ci per drum). The package used is a drum manufactured in Romania industry. The type tests carried out are described i.e. compression, penetration, free fall, leaching, safety in use (biological protection), chemical and mechanical characteristics, effect on the environment, and also the results, interpretation and conclusions from the tests. As a result of the tests, Romanian technology for treatment and packaging of radioactive wastes is considered to be in accor...