Volume 75, Issue 6

During a postulated loss-of-coolant accident, significant structural loadings on the reactor pressure vessel and its internals can be exerted due to the non-uniform and time-dependent pressure differences. The integrity of the internals must be secured, so as to ensure reactor shutdown and post-accident heat removal. Two developed computer codes to be discussed in this paper are used for the analysis of loads on the reactor pressure vessel internals. In the analyses the fluid-structure interaction effects are taken into account in the annular space between the pressure vessel and the core support barrel as well as in the upper plenum. This is a more realistic approach with respect to peak loadings and structural response frequencies.

Disposal facilities are designed to provide long-term isolation of the wastes from the human environment by means of a system of barriers, both engineered and natural. In this regard, it is required that the containment function of the multi-barrier system be evaluated by a quantitative risk analysis procedure, also called performance assessment. This entails the implementation of a Probabilistic Safety Assessment model for representing the stochastic failure behavior of the repository barriers, estimating the release rates of radionuclides into the groundwater stream and the doses from subsequent ingestion of drinking water. To this aim, a Monte Carlo simulation scheme is provided within a reliability modelling approach for the safety performance analysis of a repository, accounting also for barrier degradation processes. An application to a case study of the literature is presented to validate the approach and demonstrate its simplicity and flexibility.

The analysis of research reactor transients depends on the kinetic parameters (i.e. effective neutron generation time Λ∗ and the effective delayed neutron fraction β eff ). In this work we will calculate these parameters using the MCNP code. Effects of burn up and of Be reflector and cobalt irradiation device on kinetic parameters were also studied using the MCNP code. The results were compared with reported values. The calculations were performed for the start-up core configuration (reference core configuration) at different burn up steps with equilibrium concentration and with zero concentration of Xe, and repeated for the sixth core configuration (core configuration during the study).

Typically, reactivity initiated accidents for nuclear power plants with RBMK type reactors are classified as design basis accidents. However, for assessment of the real probability of these events, such kind of incidents should be considered more seriously and the same acceptance criteria as for operational occurrences should be applied to them. The spurious withdrawal of control rods, as limiting reactivity initiated accident (RIA), was analysed and results of the study are presented in this paper. The power distribution in fuel channels located around the affected control rods was calculated using the QUABOX/CUBBOX-HYCA code. Evaluating possible uncertainties it was found that linear power, i. e. heat generated per unit length of coolant channel, exceeds the acceptance criterion by a factor of 1.21. This paper presents the discussion how such a peak of power influences the behaviour of RBMK-1500 reactor fuel rods. For this purpose, the FEMAXI-6 code was applied. Results of the analysis show, that after increasing and subsequently decreasing of the linear power during an accident most parameters (pressure inside fuel rods, radial displacements, fuel clad gaps, etc.) return to their initial values and no phenomena occurred that would preclude the resumption of normal operation after termination of such an event.

An approach is presented that allows the computational determination of the ductility of irradiated zircaloy-4 cladding by transferring experimental results yielded under conditions atypical for normal power operation to typical loadings encountered in an usual pressurized water reactor. The transfer bases on a damage function that describes the gradual development of material separations during the cladding creep in the reactor. The numerical values of the needed equation parameters are derived from experimental results. Creep and damage evolutions are coupled leading to a simultaneous computation of both processes. The calculations indicate that the minimum strains that irradiated cladding material can accommodate are greater than 5% under pure pressure loading up to hoop stresses of 100 MPa and greater 7% when fuel swelling corresponding to a linear rod power of maximum 400 W/cm dictates the creep.

Research reactors of power greater than 20 MW are usually designed to be cooled with upward coolant flow direction inside the reactor core. This is mainly to prevent flow inversion problems following a pump coast down. However, in some designs and under certain operating conditions, flow inversion phenomenon is predicted. In the present work, the best-estimate Material Testing Reactors Thermal-Hydraulic Analysis program (MTRTHA) is used to simulate a typical MTR reactor behavior with upward cooling under a hypothetical case of loss of off-site power. The flow inversion phenomenon is predicted under certain decay heat and/or pool temperature values below the design values. The reactor simulation under loss of off-site power is performed for two cases namely; two-flap valves open and one flap-valve fails to open. The model results for the flow inversion phenomenon prediction is analyzed and a solution of the problem is suggested.

The Advanced Heavy Water Reactor (AHWR) is a pressure tube type light water cooled heavy water moderated Boiling Water Reactor (BWR) with natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the Main Heat Transport (MHT) system configuration. These multiple (four) interconnected loops influence the Steam Drum (SD) level control adversely. Such a behavior has not been reported in the open literature. The MHT configuration has been chosen based on comprehensive overall design requirements and certain Postulated Initiated Events (PIEs). This does not allow the partitioning of the Common Reactor Inlet Header (CRIH). If partitioning of the CRIH into four segments is allowed then, it will make each loop independent. Then the SD level control problems subside as the unaccounted interaction among the loops is eliminated. It has also been observed that the open loop response is stable, non-oscillatory and non-diverging for a step change in the feed flow rates. A conventional individual 3-element SD level controller cannot account for the highly coupled and interacting behaviors, of the four loops and SD levels. To overcome these interactions it is proposed to interconnect all the four steam drums in the liquid and vapor regions respectively. The influence of the interconnect configuration and the level controller are studied in detail to find a robust solution. The response obtained for unsymmetrical core power disturbances shows that the SD levels do not diverge and quickly settle to the set points assigned with SD interconnect. The proposed scheme also works well for most of the PIEs considered.

Aim of the presented study is to compare the calculated cross sections for the production of 140 Nd/ 140 Pr and 114m In/ 114 In via different reactions with incident particle energy up to 50 MeV as a part of systematic studies on particle-induced activations on metal target, theoretical calculation of production yield, calculation of target thickness requirement and suggestion for optimum reaction to produce Neodymium-140 and Indium-114m. 140 Nd/ 140 Pr and 114m In/ 114 In excitation functions via nat Ce( 3 He, xn) 140 Nd → 140 Pr, 141 Pr(p,2n) 140 Nd → 140 Pr, 141 Pr(d,3n) 140 Nd → 140 Pr, 114 Cd(p, n) 114m In → 114 In, 114 Cd(d,2n) 114m In → 114 In and 116 Cd(p,3n) 114m In → 114 In reactions were calculated by ALICE/91, ALICE/ASH (GDH Model&Hybrid Model) and TALYS-1.2 (Equilibrium&Pre-Equilibrium) codes and compared to existing data.

During the irradiation process of MoO 3 powder with in a high neutron flux, energy deposited in the powder must be released to avoid energy accumulation. The temperature increasing in the powder due to energy accumulation may cause powder phase change and under certain conditions the temperature may reach the melting point. An analytical model was developed to study the effect of gap conductance on the temperature distribution in the powder during the irradiation process. The gap conductance model was studied for helium and nitrogen gases at different pressures to obtain the optimum conditions verifying the safety in heat transfer point of view. It was found that the gap conductance is a function of gas pressure. The model provided the optimum conductivity for the gap filled with helium gas at 1 atm pressure.

The U 1 approximation is used to determine the asymptotic relaxation length (diffusion length) for one-speed neutrons in a homogeneous slab. The method is based on the series expansion of the neutron angular flux in terms of the Chebyshev polynomials of second kind and then calculating the diffusion length by applying the first order approximation to transport equation. Analytic and numerical results are obtained for the diffusion length and compared with the ones obtained from the method of separation of variables and simple diffusion theory (P 1 approximation).

The one-speed neutron transport equation which includes isotropic forward and backward scattering has been studied using the T N approximation. Critical thicknesses for different values of the reflection coefficient R and c values have been calculated. The study shows that the T N method gives nearly the same results as the P N approximation in one-dimensional geometry. The results obtained in this study are presented in tables and are compared with P N approximation results.

In this study, a simple procedure for the radiochemical determination of thorium isotopes and a robust technique for rapid preparation of uranium and thorium alpha source by electro-deposition using organic buffer solution is described. From the experiments, experience has shown that (1) TOPO is one of the best solvent for the extraction of thorium, (2) isopropyl alcohol is a suitable buffer solution for the electrodepositing of uranium; while acetone is the best choice in case of thorium, (3) the possibility of using of organic phase; containing thorium in the final chemical separation step for electrodepositing, (4) the use of only 3–5 ml of isopropyl alcohol or acetone; 3–5 mm of anode-cathode distance and relatively current 4–15 mA for electro-deposition procedure, (5) the actinides can be deposited quantitatively on a stainless steel disk within 20 to 40 min with nearly 100% of electro-deposition efficiency and the energy resolution of alpha-ray spectra of the electrode-posited sources obtained from alpha spectrometry are within the range of 37–50 KeV (FWHM).