Rbmk Ve Candu Tipi Reaktörlerde Boşluk Katsayısı Etkisinin Nötronik Açıdan İncelenmesi

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Tarih
1998-06-08
Yazarlar
Mehtap, Yalçınkaya
Süreli Yayın başlığı
Süreli Yayın ISSN
Cilt Başlığı
Yayınevi
Enerji Enstitüsü
Energy Institute
Özet
Bu çalışmada RBMK ve CANDU tipi reaktörlerde ele alınan hücre modeli için çok gruplu nötron transport kodu WIMS-AECL yardımıyla, soğutucu kaybının oluştuğu bir kaza durumu için, reaktörün kritiklik hesabı yapılarak, boşluk katsayıları nedeniyle reaktöre ithal edilecek olan reaktivite miktarı saptanmış, sonsuz çoğalma katsayısı formülündeki dört faktör incelenmiştir. Özellikle RBMK tipi reaktörlerin, soğutucu suyunun boşaltımı şeklinde modellenen bir kaza durumu esnasındaki parametreleri daha ayrıntılı incelenerek, reaktöre daha düşük miktarda reaktivite ithal edebilen malzeme konfıgürasyonu geliştirilmiştir. Bu yöndeki çalışmalarımızı dört ana başlık altında toplayabiliriz. Bunlar;. Soğutucu sıcaklığı ve yoğunluğu değişimi,. Yavaşlatıcı sıcaklığı ve yoğunluğu değişimi,. Yakıt sıcaklığı değişimi,. Farklı elementlerin zehirleme özellikleri. Merkezi tüpteki gadolinyum ya da grafitteki bor miktarı
The Chernobyl accident was the most traumatic event in the entire history of civil nuclear power. For some months we could only speculate and guess about the causes of the accident but in August 1986, the IAEA organised international conference in Vienna at which a group of Russian scientists and engineers made comprehensive presentations on what had happened. The accident of the Russian RBMK nuclear reactor, occured on 26 April 1986, was due to the three main design drawbacks: 1. The reactor had a positive void coefficient and, below 20 % power, a positive power coefficient, which made it intrinsically unstable; 2. The shutdown systems were too slow in its operation in the event; 3. There were no physical controls to prevent the staff from operating the reactor in its unstable regime or with safeguard systems seriously disabled or degraded. The accident was triggered by a turbo-generator experiment, when the reactor core contained water at just below the boiling point. When the experiment began, half of the main coolant pumps were slowed down and the flow reduction caused the water in the core to start boiling vigorously. The bubbles of steam that formed absorbed neutrons much less strongly than the water. They displaced and the number of neutrons in the core started rising. This situation increased the power of the reactor, more steam was produced and thus less neutrons were absorbed due to the phenomenon known as positive feedback. The reduction of neutron absorption caused the excess reactivity of the core to rise to the level where the chain reaction could be sustained by prompt neutrons alone, and the reactor became prompt critical. The power surge caused the fuel to heat-up, melt and disintegrate. Fragments of fuel were ejected into the surrounding water, causing steam explosions that ruptured the fuell channels and led to the pile cap being blown off. Radionuclides escaped into the atmosphere where the wind carried them to distant countries. In the last few seconds of the incident, the operators realised that the power was rising and initiated the manual trip, but the shutdown capability had been degraded by maloperation of control rods and the reactor did not shutdown immediately. Instead, a few seconds after the trip button had been pressed, the unstable reactor went prompt critical. The Soviet designed RBMK is a pressurized water reactors with individual fuel channels and using ordinary water as its coolant and graphite as its moderator. Its design is very different from most other power reactor designs as it was intended and used for both plutonium and power production. The combination of graphite moderator and water coolant is not found in the other power reactors. The reactor core, 11.8 m in diameter and 7 m high excluding the reflectors, is built up from graphite blocks perforated by vertical channels each containing a zirconium alloy pressure tube 88 mm external diameter and 4 mm thick. Each of the 1661 channels contains two fuel assemblies each 3640 mm long held together by a central tie rod, suspended from a plug at the top of the channel. The fuel assemblies consist of 1 8 pin clusters, each pin in the form of enriched (2 %) uranium dioxide pellets encased in zirconium alloy tube. About 5 % of the energy of fission is dissipated in the graphite structure as a result of slowing down of neutrons and of gamma heating. This heat is transferred to the fuel channels by conduction and radiation via a series of piston ring type graphite rings which permit good thermal contact between the pressure tube and the graphite blocks while permitting some small dimensional changes. The CANDU reactor was designed by Atomic Energy Canada Limited (AECL) as an alternative to the other reactor designs which use slightly enriched uranium (2.5 %). The CANDU fuel contains pellets of uranium dioxide with natural uranium. CANDU design consist of a horizontal calanderia which has tubes for the fuel rods and cooling water. Around these tubes is heavy water, which acts as the moderator to slow down the neutrons. In the case of the CANDU and RBMK reactor designs, refuelling may be done while the reactor is operating. Most of the world's operating power reactors have negative void coefficients. In those reactors where same water circuit acts as both moderator and coolant, excess steam generation reduces the slowing of neutrons necessary to sustain the nuclear chain reaction. This leads to a reduction in power. In some reactor designs, however, moderator and the coolant are in separate circuits, or are of different materials. In these reactors, excess steam reduces the cooling of the reactor, but as the moderator remains intact the nuclear chain reaction continues. In some of these reactors, most notably in RBMK, the neutron absorbing properties of the cooling water are a significant factor in the operating characteristics. In such cases, the reduction in neutron absorbtion as a result of steam production, and the consequent presence of extra free neutrons, enhances the chain reaction. This leads to excess power production. This excess power production causes additional heating. The additional heat raises the temperature in the cooling circuit and more steam is produced. More steam means less cooling and less neutron absorbtion, and the problem gets worse. All of this can happen very rapidly. It is very difficult to stop because it feeds itself, however if it is not stopped there will be the kind of event that happened at Chernobyl-4. In order to avoid the problems due to the positive void coefficient, there are two approaches. Either the reactor characteristics can be altered to reduce the positive void coefficient or systems can be provided that will shut the reactor down very quickly if an increase in power is detected. After the accident at Chernobyl Unit 4, the primary concern was to reduce positive void coefficient. All operating RBMK reactors, in the former Soviet Union therefore, had the following changes implemented to improve operating safety: To improve the operational reactivity margin the effective number of manual control rods was increased from 30 to 45, The installation of 80 additional absorbers in the core to inhibit operation at low power, An increase in fuel enrichement from 2 % to 2,4 % to maintain fuel burnup with the increase in neutron absorption, Scram rod insertion time cut from 18 to 12 seconds, The redesigns of control rods, The installations of a fast scram system Precautions against unauthorised access to emergency safety systems. The safety of the Soviet-designed, graphite moderated RBMK nuclear power plants is still subject to intense investigations. Among other aspects, the understanding of the neutronic behaviour of the core is essential when judging normal operating conditions and accidental scenarios. In various analyses of the Chernobyl-4 accident, the positive void effect, inherent to the original reactor core design, was identified as one of the main causes for the reactivity excursion. As a consequence, several measures have been planned and implemented to reduce the void reactivity coefficient, leading to individual designes for each of the 15 RBMK reactors now in operation. Since the experimental approach to determine the safety relevant parameters (e.g. number of absorber rods, absorber's positions, local and whole-core burn-up characteristics, short term xenon concentration, spatial perturbations, axial and radial coolant void distributions, fuel composition, power level and distribution) is very limited, an important goal is to establish adequate, reliable calculation tools. Together with pointwise 89 groups ENDF/B-V cross-section library, WIMS-AECL is able to provide high quality critical results. In this study, reactor critical facility was calculated by using multigroup neutron transport code WIMS-AECL for any lost of coolant accident in the RBMK and CANDU type reactors and the negative and positive void reactivity effect which was one of the main reasons for the catastrophic course of the Chernobyl-4 accident was identified. In addition, four factors in the infinite multiplication coefficient were examined. Especially, parameters of RBMK type reactors during lost of coolant was examined more detailed and was developed, minimum reactivity can be contributed to reactor, material configuration. Four studies can be considered as following;. Density and temperature of coolant,. Density and temperature of moderator,. Temperature of fuel,. Poison effects of different elements: Gadolinium in the central tube and boron impurities in the graphite.
Açıklama
Tez (Yüksek Lisans) -- İstanbul Teknik Üniversitesi, Enerji Enstitüsü, 1998
Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Physics, [DATE]
Anahtar kelimeler
Boşluk katsayısı, CANDU reaktörleri, Reaktörler, Void coefficient, CANDU reactors, Reactors
Alıntı