MTR tipi yakıt elemanı kullanılan araştırma reaktörlerinde soğutucu kanalı tıkanması kazası analizi

Abstract

ÖZET TR-2 gibi havuz tipi araştırma reaktörleri için göz önüne alınması gereken kazalardan birisi de soğutucu kanallarının tıkanması (Channel Blockage Accident) kazasıdır. Soğutucu kanallarından bir kısmının veya tümünün tıkanması durumunda reaktör kazadan bir süre önce durmuş olsa bile, reaktör kalbindeki artık ısının uzaklaştırılıp uzaklaştırılamayacağının hesaplanması gerekmektedir. Bu çalışmada TR-2 Araştırma Reaktörü' nün soğutucu kanallarının kısmen veya tümden tıkanması kazası, MTR tipi plaka yakıt elemanı kullanan araştırma reaktörleri için geçici ve sürekli rejimde üç boyutlu soğutucu kaybı kazası (LOCA) analizi yapabilen THEAP-I Bilgisayar Kodu kullanılarak yapılmıştır. Bu kod Uluslarası Atom Enerjisi Ajansı tarafından plaka yakıt elemanı kullanan araştırma reaktörleri'nin LOCA analizi için kabul edilmiş bir koddur. THEAP-I kodu soğutucu kanalı tıkanması kazasını analiz edebilecek şekilde modifiye edilmiş ve sonuçlar maksimum sıcaklıkların elde edildiği merkezi yakıt elemanı için sıcaklıkların çeşitli eksenel bölgelerde zamanla değişimi cinsinden verilmiştir. Kanal tıkanmasının olası bir depremin reaktör binasına vereceği hasardan dolayı bina tavanından dökülen kum, çakıl, metal gibi malzemelerle oluşacağı ve reaktörün kanallar tıkanmadan bir süre önce deprem dedektörünün depremi dedekte etmesiyle durduğu varsayılmıştır. İlk olarak reaktör kalbindeki merkezi yakıt elemanın tıkanması durumu incelenmiştir. Haftalık çalışma programı (haftada beş gün, günde altı saat çalışma şekli) ve aylık çalışma programı (bir ay sürekli çalışma) için ayrı ayrı hesaplar yapılmıştır. Sonuç olarak sadece bir aylık sürekli çalışma durumunda kalpte ergimenin oluşacağı görülmüştür. Daha sonra her iki çalışma programı için tüm kalbin tıkanması durumu incelenmiş ve ergimenin yine sadece bir aylık sürekli çalışmada oluşabileceği görülmüştür. Sonuç olarak; TR-2 Reaktörü'nde reaktör durduktan sonra oluşacak soğutucu kanalları tıkanması kazasında, kalp ergimesinin sadece sürekli çalışma programı uygulandığı sürece oluşacağı, haftalık çalışma programı gibi kesintili çalışma programlarında ise herhangi bir ergimenin oluşmayacağı görülmüştür. Bu tez çalışması, temmuz 1995'den bu yana güvenlik analizi yeniden değerlendirilmek amacıyla durdurulan TR-2 Araştırma Reaktörü'nün, kaza analizleri için gerekli olduğundan hazırlanmıştır. Güvenlik açısından, yapılan kaynama modeli yaklaşımı yüksek sıcaklıklar vermesi açısından daha konservatif olmasına rağmen, kaynama olan soğutucu kanalı bölgelerinde iki fazlı akışın gözönüne alınarak korunum denklemleri ve taşınımla ısı transferi katsayılarının yeniden tanımlanmasıyla yapılacak bir analiz daha gerçekçi olacaktır ve bu tür bir analiz tamamlayıcı bir çalışma olarak yapılabilir. IX

CHANNEL BLOCKAGE ACCIDENT ANALYSIS FOR RESEARCH REACTORS WITH MTR-TYPE FUEL ELEMENTS SUMMARY One of the postulated accidents for pool-type research and test reactors is the blockage of the coolant channels which results in loss of flow in the blocked channels. Such an accident would lead to the retention of coolant water in the coolant channels instead of natural or forced circulation of water through the channels. Once the blockage occurs, the flow rate of coolant will be decreased and main heat transfer mechanism in the blocked channels will be conduction that could result in less heat transfer rate when compared to convective heat transfer rate. It is necessary to demonstrate that the residual decay heat can be removed without the hazard of core melting in such a channel blockage accident event. It is the purpose of this study to investigate the feasibility of removing the residual decay heat from a pool-type research reactor core after a channel blockage accident event following an earthquake and to identify the principal factors involved in cooling process. Nuclear plants are designed to withstand the ground motion caused by the most severe earthquake that is likely to be experienced. From historical records of seismic events in the plant vicinity, the earthquake that would be expected to produce the largest ground motion at the reactor site is predicted. This is called the `safe shutdown earthquake.` Analysis must show that the reactor can be tripped and the engineered safety features will function properly if such an earthquake should occur. In the current study the following accident scenario is considered to analyze the partial or entire coolant channel blockage of TR-2 Reactor following a severe earthquake. TR-2 Reactor plant is structurally designed to withstand a maximum horizontal acceleration of 0.4g as the result of the safe shutdown earthquake. It isassumed that the reactor is operating at full power and is essentially instantaneously shut down when the seismic ground motion is detected by the earthquake detector. It is further assumed that the reactor building will be damaged as a result of the earthquake and some granular materials such as sand, soil will fall down to the open pool surface causing blockage of some coolant channels of the reactor core. The blockage of any channel is considered fully but not partially to simplify the analysis. Otherwise, the heat transfer and conservation equations for blocked channels should be written as for porous media which complicates the modeling too much. In the present modeling, the coolant flow is not permitted by means of natural circulation of water through blocked channels. From the time when the reactor is shutdown to the time when the coolant channel is blocked by the fallen particles from the roof of the reactor building, the removal of decay heat from the reactor core is considered by means of natural convection of coolant water. The time at which the channels become blocked is considered the `start` of the Channel Blockage Accident, i.e., t=0 in the computer modeling used in the analysis. On the other hand, after the blockage, main decay heat removal mechanism of the core is considered as conduction in the blocked channels and natural convection in the unblocked channels. Conduction and convection to the neighboring elements or radiation heat transfer to the outer surfaces from the reactor core is also considered in the modeling. To analyze the above hypothetical accident event, it was decided to use THEAP-I computer code, a single phase transient 3-D structure/ 1-D flow thermal hydraulics code developed with the aim to contribute mainly to the safety analysis of the open pool research reactors. THEAP-I can be used to examine a set of thermally interacting heated/cooled channels which are kept together through one support plate located at the elevation. It is considered that whole bundle and the support plate are totally immersed into water or air and the thermal interaction among channels or/and the support plate occurs through small gaps of finite (or zero) length. XIThe channel structure is cooled through an internal flow stream naturally or forcibly convected. At the first stage of the problem modeling, THEAP-I was modified to correspond the current accident scenario. Initially, the blocked channels are defined by means of setting flow rates to `zero` for the selected channels in the subroutine of THEAP-I which calculates the natural flow rates (NATFLO). To introduce the blocked channels to the code a specific hydraulics diameter value is assigned for blocked channels in the problem input data so that whenever hydraulics diameter matches to assigned value, calculations will be done for the blocked channels. Then, to evaluate the equivalent heat transfer coefficients for the blocked channels the convection term is dismissed in the subroutine (SCOOLD) calculating equivalent heat transfer coefficients for vertical channels. The second step was to complete the physical modeling by modifications in the code in order to calculate boiling heat transfer coefficients for the channel sub- regions where the boiling temperature is reached. For THEAP-I is a single phase thermal hydraulics code a rigorous approximation, which is more conservative than the boiling itself, was done to simulate the boiling, To do this the bubbles in the boiling regions are treated as air. Because of the worse heat transfer coefficients of air than vapor at the same tempertaure this assumption leads to higher temperature estimations and in view of safety aspects is more conservative than the actual situation. Following the code modifications, centered fuel element of TR-2 Reactor was considered blocked for the first run of the code. Since the reactor's operating schedule has an essential role in the calculation of the core decay heat two operating schedules was considered : 1. weekly operating schedule (operation of five day per week, six hours a day), 2. one month continuous operation. The results of the analysis for each operating schedules are given in terms of temperature field distribution of the central fuel element as a function of the time. XIIThe worst case, blockage of the whole core channels, was considered for both operating schedules too. The results of analysis for each schedule are given in terms of temperature fields distribution of the centered fuel element as a function of the time. It should also be noted that one of the important program input parameters to be determined was the time `zero`, the time passed after earthquake until blockage of the channels occurred. Because, selection of different `zero` times will result in different initial decay heat and initial temperature fields of the core elements at the time of blockage. Successive runs showed that the behavior of the thermal hydraulics system was purely sensitive to the input parameter `zero` time in case, it is in the orders of up to a few minutes. So, a reasonable value of 10 seconds assigned for `zero` time in all of the above four runs. All of the analysis results figured out the fact that the core melting was inevitable in case of an uninterrupted operation (continuous operation) preceding a channel blockage accident of the TR-2 Reactor. Such a result will even be met if the blockage occurs only in a single fuel element. But this is not the case for the reactor's routine operating schedule. The reactor's operation scheme had been weekly operating schedule until it was ceased in the mid-summer of 1 995 in order to re-evaluate safety aspects of it. The analysis also shows that core melting will not occur if the reactor is operated in weekly schedule which is not a continuous operating schedule that yields less decay heat rate compared to continuous operation before the accident. As a result, we could easily say that in routine operation (weekly operation) core melting will not occur following a channel blockage accident caused by an earthquake even if all coolant channels are blocked. The essential contribution of this study to the Safety Analysis Report of TR-2 Reactor has been compensating the lack of `accident analysis` section. Further amendments could be made by taking into consideration of boiling heat transfer coefficients in the coolant channel regions where boiling occurs. XIIIThe results figured out the fact that a detailed channel bloackage accident analysis considering two phase heat transfer mechanism must be made if TR-2 Reactor is planned to be operated in a continuous operating program. The reevaluation of the accident would be necessary because of the higher temperature estimations of the current study. Otherwise, the current results indicate the risk of core meltdown in case of a channel blockage accident. XIV