الفهرس Only 14 pages are availabe for public view
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Abstract
Radioactive release from nuclear reactors under normal and accidental conditions is one of important issues in reactor safety assessment. The release limits should be controlled and stated in the preliminary safety analysis report (PSAR) and final safety analysis report (FSAR). In the nuclear safety analysis, quantitative evaluation of the source term under routine operation and accidental conditions should be given. Going back to the main productive source of fission products, i.e. the fuel element. the amount of fission products existing in a given reactor core at a given time during normal operation can be accurately estimated from the operation history of the core, as well as, accomponing releases. The fraction of fission products that will be released under a given set of circumstances (accidental releases) cannot be accurately estimated in an easy way. The releases can be from the fuel to primary coolant, from primary coolant to secondary coolant, from secondary coolant to
containment, and from containment to environment.
The present study considers the releases from fuel to coolant. The source term for this releases is the nuclides generated in the fuel matrix due to the fission process, it may be called radionuclide inventories in the fuel, it is composed of actinides, fission products, and other radionuclides. The formations of these nuclides depend on different factors such as reactor design, materials used in the reactor, operating power level, operating history, and fuel burn-up. Different models, methods, and experimental work, have been used in determining the possible fission products materials that may be released at fuel-clad interface, reactor coolant, outside reactor and to the surrounding environment of an operating reactor. Upon determining the routine releases in coolant of an operating reactor, one may assess the efficency of filtering systems, normal doses to workers, and examine if these limts fufilled the safety requirements imposed by regulatory body. Also, the releases from a virtual desgin base accident have been determined, aiming to verify its compliance with the safety regulations.
The study considers a case of a MTR-type reactor (Material Test and Research reactor).
Models and codes are considered as valuable documents. For the source term calculations, all radionuclides produced as sustained fission continues, has been conducted using ORlGEN2 code.
Although the code does not have specific nuclear data library for MTR-type reactor, a closer data pertained to similar flux pattern of PWR are utilized. The output of the code has been validated using a simple production and decay model against published data. The time variable in running ORIGEN2 was taken as both actually operated real time with average power days and full power days concepts. This is practically accepted especially if the species is a long lived radionuclide. Reduction factors that represent releases at fuel - clad interface, cladding defect and radioactive decay have been studied. Contribution of all these factors with the amount produced in the fuel gives an estimator to the concentration of any selected species in the reactor coolant. The cladding defect under accidental conditions was studied and its significance against the severity of the accident has been assessed for validating mathematical release models as well as source term computer codes.