Volume 77, Issue 5

The trends towards higher burnups, longer in-core fuel residence time, and the development of new cladding materials, have led to the need to re-assess the safety criteria for loss-of-coolant accidents. The HALDEN IFA-650 series of experiments are integral, single-pin, in-pile tests on the fuel behavior undergoing conditions of a loss-of-coolant accident over pre-determined fuel rod linear heat rate, and under thermally-controlled structural environment. Fresh and irradiated fuel pins from commercial reactors were tested. The present work seeks to assess the capability of a new (best-estimate) thermal-hydraulics code TRACE (version 5.0), to adequately predict the thermal-hydraulic impact of a loss-of-coolant accident on fuel rods. Several tests have been simulated over a range of fuel rod power and surrounding structure temperatures. The tests initiated from conditions of single-phase liquid natural circulation operating at 7MPa with subsequent venting to near-atmospheric pressure.

This paper will provide results of an uncertainty and sensitivity study in order to calculate parameters of safety related importance like the fuel centerline temperature, the cladding temperature and the fuel assembly pressure drop of a lead-alloy cooled fast system. Applying best practice guidelines, a list of uncertain parameters has been identified. The considered parameter variations are based on the experience gained during fabrication and operation of former and existing liquid metal cooled fast systems as well as on experimental results and on engineering judgment.

Deterministic safety analysis is an important tool for confirming the adequacy of provisions within the defense-in-depth concept for the safety of nuclear power plants. One of the important design basis events is considered to be a complete double-ended guillotine rupture i.e. Loss of Coolant Accident (LOCA) of largest and coldest pipe in the primary coolant circuit. The present work deals with this scenario for an Indian PHWR. It highlights the identification of critical break size leading to the maximum clad temperature using the best estimate code RELAP5. Further, important parameters affecting the clad temperature are described along with results of initial sensitivity studies to select dominant uncertain parameters. For the uncertainty propagation, Latin Hypercube Sampling (LHS) is used instead of simple random sampling for Monte-Carlo simulation. The inherent characteristic of LHS is to reduce the required runs for Monte-Carlo simulation to manageable order for the current computing capability. The 95 th percentile value of peak cladding temperature (PCT) is obtained by the method and compared with acceptance criteria.

Modern theoretical methods for analysing the stability behaviour of Boiling Water Reactors (BWRs) are relatively reliable. The analysis is performed by comprehensive validated system codes comprising 3D core models and one-dimensional thermal-hydraulic parallel channel models in the frequency (linearized models) or time domain. Nevertheless the spontaneous emergence of stable or unstable periodic orbits as solutions of the coupled nonlinear differential equations determining the stability properties of the coupled thermal-hydraulic and neutron kinetic (highly) nonlinear BWR system is a surprising phenomenon, and it is worth thinking about the mathematical background controlling such behaviour. In particular the coexistence of different types of solutions, such as the coexistence of unstable limit cycles and stable fixed points, are states of stability, not all nuclear engineers are familiar with. Hence the part I of this paper is devoted to the mathematical background of linear and nonlinear stability analysis and introduces a novel efficient approach to treat the nonlinear BWR stability behaviour with both system codes and so-called (advanced) reduced order models (ROMs). The efficiency of this approach, called the RAM-ROM method, will be demonstrated by some results of stability analyses for different power plants.

The pebble bed reactor is one of the most promising concepts for (V)HTR. Nevertheless recent re-evaluation of AVR and THTR results emphasizes once more that claimed advantages strongly depend on the properties of the fuel elements and their behavior under operational conditions. Additionally safeguards, waste management and disposal aspects gain increasing importance today. The conventional uncoated graphite pebbles meet only inadequately the requirements of Generation IV facilities. Since long experts agree, that corrosion-resistant pebbles with high retention capability for fission products would considerably improve the chances of the pebble bed reactor concept. With laser beam joining of ceramics the key technology is now available for the silicon carbide (SiC) encapsulation of (V)HTR components. The envisaged innovative fuel element consists of a robust SiC hollow sphere filled with moderator and TRISO coated particles. The positive assessment according to Gen IV criteria should justify necessary R&D efforts to obtain qualified fuel elements and demonstrate their superiority under operational conditions.

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.

This paper presents a neural network based fault diagnosing approach which allows dynamic crack and leaks fault identification. The method utilizes the Principal Component Analysis (PCA) technique to reduce the problem dimension. Such a dimension reduction approach leads to faster diagnosing and allows a better graphic presentation of the results. To show the effectiveness of the proposed approach, two methodologies are used to train the neural network (NN). At first, a training matrix composed of 14 variables is used to train a Multilayer Perceptron neural network (MLP) with Resilient Backpropagation (RBP) algorithm. Employing the proposed method, a more accurate and simpler network is designed where the input size is reduced from 14 to 6 variables for training the NN. In short, the application of PCA highly reduces the network topology and allows employing more efficient training algorithms. The accuracy, generalization ability, and reliability of the designed networks are verified using 10 simulated events data from a VVER-1000 simulation using DINAMIKA-97 code. Noise is added to the data to evaluate the robustness of the method and the method again shows to be effective and powerful.

Analytical solution methods for the neutron noise in a one-dimensional multi-region system in two-group theory, which have so far been based on the adjoint function technique, are extended here to using the forward Green's function technique. The forward Green's functions were calculated analytically for a noise source in a core surrounded by reflector regions at both sides. It is shown that with symbolic computation methods, the forward Green's function can be used for the calculation of the space- and frequency-dependent noise in the first order approximation for arbitrary noise sources which have an analytical representation. The properties of the induced neutron noise were investigated for vibrations of both absorbers and fuel assemblies, with two representations of the noise sources: a point-like source which corresponds to the vibrations of a fuel rod, and a finite width source which corresponds to vibrations of a fuel assembly. The contributions of the components induced by the fluctuations of the various types of macroscopic cross sections in the total noise are also discussed and the information content of the noise in the fast group is explored for the identification of fuel assembly vibrations.

The first order approximations of the Legendre and Chebyshev polynomials are applied to neutron transport equation to calculate the diffusion length for one-speed neutrons in a uniform homogeneous slab. The Anli-Güngör (AG) phase function is used for the scattering kernel. Analytic and numerical results are obtained for the diffusion length of the neutrons in terms of scattering parameters of AG phase function and cross-section parameters. The numerical results for certain scattering parameters of AG phase function and selected values of cross-section parameter from first order approximations of both Legendre and Chebyshev polynomials are tabulated in the tables for comparison.

Gold-198 (β max =0.96MeV (98.6%), γ max =0.412MeV (95.5%) and T 1/2 =2.7 days) is a well-known therapeutic beta emitter in the field of nuclear medicine, and is being used for the treatment of many different cancers. In the present study, the Bremsstrahlung exposure induced by 198 Au in different human tissues, DNA and RNA has been calculated. The specific Bremsstrahlung constant (Γ Br ), Probability of energy loss by beta during Bremsstrahlung emission (P Br ) and Bremsstrahlung activity (A release ) Br were estimated. We strongly recommend these parameters should be considered in absorbed dose calculations of radionuclide therapy via 198 Au.

This study aims to determine AAPM TG-60 dosimetric quantities in regions surrounding the 103 Pd stent wall by MCNP5 Monte Carlo code. The Palmaz-Schatz stent was modeled by a hollow cylinder of 17.89mm length (2mm diameter) with net surface very similar to real stent. The Dose deposited per photon (Gy), relative dose, Anisotropy function, F(r,θ), and radial dose function, g L (r), were described at AAPM TG-60 protocol were generated from these values and listed in tabular format. For benchmarking, the relative dose values were verified with TG-43 and EGS4 code results at identical conditions, relative to the radial distances from surface of the stent. There were noticeable results. These physical dosimetric parameters can be used in future treatment planning systems for IVBT.

This work presents a study of natural radioactivity levels in coal and its combustion residues (fly ash and slag) used in the houses in Black Sea Region, Turkey. Coal, fly ash and slag samples were provided from different locations of the region and analyzed by gamma spectroscopy using a high-purity germanium detector (HPGe). Also, chemical analyses of these samples were carried out using energy dispersive x-ray fluorescence spectrometer. The mean 226 Ra activity concentrations in coal, slag and fly ash were measured as 83, 99 and 38Bqkg –1 , respectively. The mean 232 Th activity concentrations in coal, slag and fly ash were measured as 108, 113 and 50Bqkg –1 , respectively. The mean 40 K activity concentrations in coal, slag and fly ash were found to be 366, 381 and 204Bqkg –1 , respectively. The potential radiological hazards associated to these materials were evaluated by calculating the radium equivalent activity (Ra eq ), the air absorbed gamma dose rate (D), the annual effective dose rate (AED), the external hazard index (H ex ) and internal hazard index (H in ) and compared with the internationally accepted or reference values. The mean Ra eq values of the coal, fly ash and slag samples were lower than the recommended maximum values 370Bqkg –1 by the Organization for Economic Cooperation and Development (OECD). The overall mean outdoor terrestrial gamma air absorbed dose rate in coal, fly ash and slag samples are 119, 129 and 62nGyh –1 and the corresponding outdoor annual effective doses are 0.60, 0.32 and 0.64mSvy –1 , which is higher than the worldwide average (0.07mSvy –1 ), respectively. Moreover, the enrichment factors relative to the input coal are calculated for the radionuclide contents observed. Calculated enrichment factor values for 226 Ra and 232 Th were found 1.14 and 1.01, respectively.