Band 76, Heft 4

The reactor pressure vessels (RPV) of boiling water reactors (BWR) belonging to the product line '69 have unusually designed heads. The spherical cap-shaped bottom head of the vessel is welded directly to the support flange of the lower shell course. This unusual construction has led repeatedly to controversial discussions concerning the limits and admissibility of stress intensities arising in the junction of the bottom head to the cylindrical shell. In the present paper, stress analyses for the design conditions are performed with the finite element method in order to determine and categorize the occurring stresses. The procedure of stress classification in accordance with the guidelines of German KTA 3201.2 and Section III of the ASME Code (Subsection NB) is described and subsequently demonstrated by the example of a typical BWR vessel. The accomplished investigations yield allowable stress intensities in the considered area. Additionally, limit load analyses are carried out to verify the obtained results. Complementary studies, performed for a torispherical head, prove that the determined maximum peak stresses in the junction between the bottom head and the cylindrical shell are not unusual also for pressure vessels with regular bottom head constructions.

We investigate steam condensation induced water hammer (CIWH) phenomena and present experimental and theoretical results. Some of the experiments were performed in the PMK-2 facility, which is a full-pressure thermalhydraulic model of the nuclear power plant of VVER-440/312 type and located in the Atomic Energy Research Institute Budapest, Hungary. Other experiments were done in the ROSA facility in Japan. On the theoretical side CIWH is studied and analyzed with the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. The applied two-fluid model shows some similarities to RELAP5 which is widely used in the nuclear industry to simulate nuclear power plant accidents. New features are the existence of multiple, independent CIWH pressure peaks both in experiments and in simulations. Experimentally measured and theoretically calculated CIWH pressure peaks are in qualitative agreement. However, the computational results are very sensitive against flow velocity.

A new nuclear power reactor under design study is a vertical pressure tube type boiling light water cooled and heavy water moderated. One of the passive design features of this reactor is the heat removal through natural circulation of primary coolant at all power level with no primary coolant pumps. Nuclear plants are mainly base load units, but the proposed plant with various advance features has to operate in load following mode i.e. Reactor follows Turbine (in a limited range). In this mode, any alteration in turbine load results in the steam pressure change. The steam pressure error is fed to the Reactor Regulating System (RRS), which changes the reactor power to control the system pressure. To study this mode of plant operation, a plant simulation model with the feedbacks from various controllers has been developed using the RELAP5 code. This integrated plant model has been used for simulating the load-varying scenario for a change in plant load. All the process dynamics, modeling, design verification and performance issues are discussed in this paper.

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.

The paper presents the experimental data of the local heat transfer in the fuel rod assembly with honeycomb type spacer grids. It was established experimentally that the separation of the boundary layer behind the spacer grid increases the interchannel coefficient of transverse mixing of flow, and therefore, the local heat transfer. If the distance between the spacer grids is relatively small, they influence one another through the redistribution of mass flow in elementary channels. The local heat transfer behind the spacer grid varies with a distance from it. The local heat transfer can be determined within ±3% using similar equations for any elementary cell.

The main objective of the PHEBUS FPT1 experiment was to study the release of fission products and their subsequent transport and deposition in the primary circuit and the containment under the conditions representative of a severe accident of a Pressurised Water Reactor. The FPT1 test was divided into the bundle degradation, aerosol, washing and chemistry phases. The objective of this paper is related to the best estimate analysis of the bundle degradation phase using ASTEC code ICARE module. GRS (Germany) best estimate method was used for the simulation of FPT1 test by employing SUSA 3.5 package. The performed best estimate analysis showed that ICARE module is capable to model the main severe accidents phenomena in the fuel bundle, heating and melting, taking into account the used physical and code parameters uncertainties.

In the complex thermal hydraulics codes developed for fire, reactor and containment safety the junctions in the multi-compartment geometries are often modeled as uni-directional junctions and some construct of flow coefficient. However, certain large size junctions are known to depict bi-directional flow behaviour under specific circumstances. The CFD based computer code FDS was used for an earlier reported study of fire in an enclosure on the bidirectional flow behaviour in present of a wall opening. Numerical simulation is directed to monitor the entrainment of the fresh air from outside to the fire compartment and resulting plume deflection due to presence of a big opening. The paper presents the details of the analysis, the results obtained, and comparison with the reported experimental data in terms of plume deflection, entrainment. Detailed investigations have been carried out to understand the bi-directionality of a junction by analyzing studying the outgoing hot air flow and incoming cold air.

Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k ∞,0 = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are ∼2.7, 8.4, and 15 years and with burn ups of ∼72000, 222000 and 366000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced.

Silver electrodeposition on gold layer was carried out by the alkaline plating baths to produce cadmium-109 via nat Ag(p, n) 109 Cd nuclear reaction. Gold was electrodeposited on copper backing in the beginning; the bath content consisted of 17.7 gl −1 KCN, 6.6 gl −1 Au, 6.6 gl −1 K 2 CO 3 and 3.3 gl −1 Na 2 CO 3 while acidity and temperature of the bath adjusted at 10 and 45°C respectively. A DC current density of ca 2.08 mA · cm −2 was used to electroplate gold. A gold layer of 63 μm thickness with suitable morphology was obtained after the electrodeposition. Silver was electrodeposited on the gold layer with 100% efficiency using a cyanide bath. The silver target was irradiated with 15 MeV proton beam and current of 150 μA; the 109 Cd production yield was 2.0 μCi/μA · h (0.074 MBq/μA · h). The target material was dissolved by 14 M HNO 3 . Cadmium-109 in the dissolved target solution was separated from silver by evaporation process. Cadmium recovered with more than 88% efficiency involved non-detected silver impurity.

Analysis has been done for accidental release of radioactive gases from high and low enriched fuel of miniature neutron source reactors (MNSR). Calculations have been performed for existing high enriched uranium (HEU) fuelled core and prospected low enriched uranium (LEU) fuelled core. Composition of existing HEU fuel is UAl 4 -Al while prospected LEU fuel is UO 2 . The ORIGEN2 code has been used for burn up and source term calculations. The “Hotspot” version 2.06 has been employed for radiological dispersion analysis to model the release of radioactivity in case of accident. Site specific meteorological data has been employed to determine the dose values for different stability classes as a function of downwind distance. In case of a radiological accident releasing radioactive gases at ground level, the maximum Total Effective Dose Equivalent (TEDE) is about 0.26 Sv for the most prevailing meteorological conditions at the site. There were no appreciable differences found in doses for HEU and LEU fuel.

In this study, a simple procedure for the determination of natural uranium, thorium and radium isotopes in water and soil samples by alpha spectroscopy is described. This procedure allows a sequential extraction polonium, uranium, thorium and radium radionuclides from the same sample in two to three days. It was tested and validated with the analysis of certified reference materials from the IAEA.