Volume 75, Issue 5

Online condition monitoring or online monitoring (OLM) uses data acquired while a nuclear power is operating to continuously assess the health of the plant's sensors, processes, and equipment; to measure the dynamic performance of the plant's process instrumentation; to verify in-situ the calibration of the process instrumentation; to detect blockages, voids, and leaks in the pressure sensing lines; to identify core flow anomalies; to extend the life of neutron detectors and other sensors; and to measure the vibration of reactor internals. Both the steady-state or DC output of plant sensors and their AC signal or noise output can be used to assess sensor health, depending on whether the application is monitoring a rapidly changing (e.g., core barrel motion) or a slowly changing (e.g., sensor calibration) process. The author has designed, developed, installed, and tested OLM systems (comprised of software/hardware-based data acquisition and data processing modules) that integrate low-frequency (1 mHz to 1 Hz) techniques such as RTD cross-calibration, pressure transmitter calibration monitoring, and equipment condition monitoring and high-frequency (1 Hz to 1 kHz) techniques such as the noise analysis technique. The author has demonstrated the noise analysis technique's effectiveness for measuring the dynamic response-time of pressure transmitters and their sensing lines; for performing predictive maintenance on reactor internals; for detecting core flow anomalies; and for extending neutron detector life. Integrated online condition monitoring systems can combine AC and DC data acquisition and signal processing techniques, can use data from existing process sensors during all modes of plant operation, including startup, normal operation, and shutdown; can be retrofitted into existing PWRs, BWRs, and other reactor types and will be integrated into next-generation plants.

Criticality and burn up behavior of the Fixed Bed Nuclear Reactor (FBNR) are investigated for the mixed fuel 233 UO 2 /ThO 2 as an alternative to low enriched 235 UO 2 fuel. CERMET fuel with a zirconium matrix and cladding has been used throughout the study. The main results of the study can be summarized as follows: Reactor criticality is already achieved by ∼2% 233 UO 2 with the mixed 233 UO 2 /ThO 2 fuel. At higher 233 U fractions, reactor criticality rises rapidly and exceeds k eff > 1.5 already by 9% 233 UO 2 . With 100% 233 UO 2 , start up criticality can reach k eff = 2.0975. Time dependent reactor criticality k eff and fuel burn up have been investigated for two different mixed fuel 233 UO 2 /ThO 2 compositions, namely: 4% 233 UO 2 + 96% ThO 2 for a reactor power of 40 MW el (120 MW th ) and 9% 233 UO 2 + 91% ThO 2 for a reactor power of 70 MW el (210 MW th ). Sufficient reactor criticality (k eff > 1.06) for continuous operation without fuel change can be sustained during ∼5 and 12 years with 4% and 9% 233 UO 2 fractions in the mixed fuel, leading to burn ups of ∼36000 and >105000MWD/t, respectively. Thorium based fuel produces no prolific uranium. Plutonium production remains negligible.

The hot channel in a typical Material Testing Reactor (MTR) is subjected to 3D simulation. Because of the existence of similarity planes, only a quarter of the hot channel including meat thickness, clad, and coolant channel is considered for CFD analysis using the FLUENT code. For the simulation, steady state normal operation regime at the reactor nominal power is assumed. In order to build confidence in our modeling approach, the results obtained in this work are compared with those obtained from the one-dimensional simulation code, MTRTHA. That is, modified variables were generated in order to match those obtained by MTRTHA and to allow comparisons. Quite good agreement is generally observed, however, the maximum clad surface temperature predicted by the 3D calculations, located at the clad mid-width, is higher than the 1D prediction by about 8°C but still below the onset of subcooled boiling by adequate safety margin. The results show quite interesting 3D patterns in both the flow field and the heat transfer. Temperature profiles, velocity profiles and contours are all presented to highlight the essential 3D features of this system.

Nuclear fusion has the potential to cover a major part of mankind's energy needs under the conditions of ever-increasing population and energy consumption. Wendelstein 7-X (W7-X) is a fusion experiment currently being built in Greifswald, which shall demonstrate the capability of the stellarator concept on the way to a fusion power plant. This paper presents analyses of processes in the target cooling system with the codes RELAP5 and ASTEC. The loss-of-coolant accident (LOCA) during the “baking” operation mode of W7-X (i.e. under the most critical conditions because of the high temperature and pressure applied) is analysed. The response of the W7-X “baking” and main cooling circuits and the amount of coolant discharged into the plasma vessel are investigated.

85 Sr, having a half-life of 64.8 d, is an important radionuclide in nuclear medicine diagnostic and endotherapy. The ALICE/ASH code was used to calculate excitation functions for proton, alpha and deuteron induced on various targets that lead to the production of 85 Sr radioisotopes using intermediate energy accelerators. Recommended thickness of the targets according to SRIM code was premeditated. The application of those data, particularly in the calculation of integral yields, is discussed and theoretical integral yields for different reactions were computed. To consider precision of ALICE/ASH code calculations, experimental excitation functions in several decay channels was compared with ALICE/ASH code data. The 85 Rb(p, n) 85 Sr process was determined as most interesting one due to radionuclidic purity. The ALICE/ASH code predicts a maximum cross-section of about 798 mb at 11 MeV for this reaction.

The purpose of this work is to study the feasibility of the Thorium/Uranium fuel cycle in heterogeneous Pressurized Water Reactors (PWR) core design. This paper focuses on the neutronic and burn-up analysis of the Thorium/Uranium fuel using the computer codes MCNPX and WIMS. The design is based on the Whole Assembly Seed and Blanket (WASB) concept, in which the individual seed (Uranium) and blanket (Thorium-Uranium) units occupy one full-size PWR assembly in a checkerboard core configuration. The results of the present models were compared with the solution of benchmark problems and satisfactory agreement was found.

Corrosion of steam generator heat transfer tubes (SGHTT) is one of the important problems which affect safety operation of nuclear power plants (NPP), and the hazard of pitting corrosion of heat transfer tubes is the most serious. With an acoustic emission device, the signals during a corrosion test on SGHTT were collected and analyzed, and the corrosion points in the tubes were located accurately. The results show that pitting corrosion of heat transfer tubes has passed through three periods in its development: expansion phase, stationary phase and rapid developing phase. The corrosion damage of HTT can be found earlier with acoustic emission than any other non-destructive testing methods. Acoustic emission can be used for on-line and real-time monitoring of the safety and operation of the steam generator and has therefore a great significance.

The transport of one-speed neutrons has been studied in isotropically scattering slab using ultraspherical polynomial approximation whose sub-cases are spherical harmonics approximation and Chebyshev polynomial approximation of first and second kind with reflective boundaries. An integral equation for the total flux has been derived and solved numerically for different values of c and reflection coefficients. In the solution, Marshak type vacuum boundary conditions were used. The critical thicknesses obtained by the Chebyshev polynomial method are in good agreement with literature values. Computations were made by using the T N , U N and P N approximations for comparison.

A modified version of the U N method is used to solve the critical slab problem in the case of a combination of forward and backward scattering with isotropic scattering in a uniform finite slab. The Marshak boundary condition is used for the calculation of the critical slab thicknesses and it is shown that the method gives accurate results in one-dimensional geometry and converges rapidly with easily executable equations. The presented numerical results are compared with the results available in literature.

In this study the half-space albedo problem has been investigated for highly anisotropic scattering. For this highly anisotropic scattering the İnönü scattering has been added to the pure-quadratic anisotropic scattering. The general solution of the one-speed neutron transport equation can be written as linear combination of the solutions which involve these scatterings. Hence, Case's eigenfunctions and the related semi-analytical methods can be used to get the albedo calculations. Some selected numerical data have been tabulated. To compare these results with the İnönü scattering values the linear interpolation technique has been used. Successful results have been achieved.

A thick layer of Yttrium oxide was deposited on cupper substrate via two special sedimentation methods for the production of 89 Zr. For deposition of Y 2 O 3 on cupper substrate, 435 mg of Y 2 O 3 , 152.25 mg of ethyl cellulose and 4 ml of acetone were used to prepare a Y 2 O 3 layer of 11.69 cm 2 . Also, 435 mg of Y 2 O 3 , 174 mg of methyl cellulose and 7.25 ml of water were used. The targets were checked by SEM and thermal shock test. The deposited target was irradiated for 20 min at a current of 20 μA and a proton beam of 15 MeV.