Fission products basis accident

Ionizing Radiation Sources and Biological Effects. New York, United Nations. U.S. Nuclear Regulatory Commission (1981) Technical basis for estimating fission product behaviour during LWR accidents. Report NUREG-0772, Washington, D.C. 

Most fission products (including cesium and iodine) and all actinides escaping the solid AHTR fuel are soluble in the molten salt and will remain in the molten salt at very high temperatures. Fluoride salts were chosen for the liquid-fueled molten-salt reactor because actinides and fission products dissolve in the molten salt at very high temperatures. This same characteristic applies to the AHTR and provides the reactor with a second, independent beyond-design-basis-accident mitigation system to prevent radionuclide release to the environment. 

Design Basis Accidents Fission product retention Coated fuel particles Pressure essel unit Passive Passive/Active - Inherently safe fission product retention - Automatic closure fail-safe isolation valves - Unfiltered release below permissible limits (normal and disturbed operation) - Further reduetion by filtered venting of reactor building in case of disturbed operation... 

Water Air Ingress Severe Accidents Fission product retention Cooling water systems Same systems and design features as tor Design Basis Accidents Active/Passive Passive/Achve - Automatic isolation - valves - Low-temperature, low-pressure coolers - See design Basis Accidents - No accident with fuel heatmg above permissible limits... 

Fission product retention Same systems and design features as for design basis accidents Passive/active Same safety features as for design basis accidents... 

The report does not aim at exhaustiveness, nor at direct applicability to severe accident management situations. Its aim is to give a picture of what is known in the area of fission product sources and how this knowledge can be put to use to mitigate them. Implementation has to be developed on a plant-by-plant basis, taking account of plant specifics. This is the task of the utilities. 

Generic Safety Issue (GSI) 82 in NUREG-0933 (Reference 1), addresses the potential for a beyond-design-basis accident in which the water is drained out of the spent fuel pool, allowing the Zircaloy fuel cladding to ignite and thus release fission products from the spent fuel to the atmosphere. (The spent fuel pool is usually located outside the primary containment.)... 

For BWRs, the radiologically representative design basis accidents have not been defined as precisely as for PWRs. However, the possible events are phenomenologically quite similar (with the exception of the steam generator tube rupture accident, which does not apply for BWRs) and there are no significant differences to be expected in radionuclide behavior. Consequently, equivalent requirements are applied to the design of both PWR and BWR plants. This means that in most cases the radiochemistry principles discussed in what follows will apply mutatis mutandis to boiling water reactors as well. In the event of a loss-of-coolant accident, the BWR pressure suppression pool represents an effective retention system the behavior of the fission products in this pool will be discussed in Section 7.3.2.4. 

As was discussed above, the release of fission products from the fuel rods which failed during the burst, refill and reflood phases of the loss-of-coolant design basis accident is limited to their gap inventories due to the comparatively low maximum... 

The consequences of a category 1 accident do not exceed those of a design basis loss-of-coolant accident (see Section 6.2.1.). In contrast, in the course of accidents belonging to categories 2 and 3, the fuel is damaged by overheating to the point of allowing substantial escape of fission products to the primary circuit and to the containment. 

The low-pressure scenario which is initiated by a large-sized break in the primary circuit, that is an event similar to that of the loss-of-coolant design basis accident described in Section 6.2.1.. In the severe accident scenario it is additionally postulated that, after the action of the accumulators and the borated water storage tanks, the sump water recirculation pumps will fail to operate. Thus, the decay heat cannot be removed from the reactor core vnth the consequence that the water volume present inside the reactor pressure vessel (RPV) begins to boil off at about atmospheric pressure. The AB sequence of WASH-1400 describes such a large-break scenario. In this low-pressure scenario, the treatment of fission product behavior inside the primary circuit is comparatively simple the probability of occurrence of such an accident, however, is extremely small. 

Similar experimental conditions to those in the Sascha experiments, that is a flowing steam-hydrogen atmosphere at atmospheric pressure, were used in the ORNL high temperature tests, the principal aim of which was to determine the impact of specific accident conditions on the behavior of fuel and fission products. In the early ORNL high-temperature experiments (HT series), fuel rod segments fabricated of power-reactor irradiated fuel and encapsulated in Zircaloy cladding were heated for a short time (a few minutes) to temperatures up to 1900 K in atmospheres of various composition. The results obtained in these tests were the main basis for the assessment of fission product release from overheated fuels made in the NUREG-0772 report (US NRC, 1981). 

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