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[en] The irradiated FM(Fission-Molly) target is unloaded from the irradiation hole during normal operation, and then cooled down in the reactor pool for a certain period of time. Therefore, it is necessary to identify the minimum decay time needed to cool down FM target sufficiently by natural convection. In the present work, numerical simulations are performed to predict cooling capability of a FM target cooled by natural convection using commercial computational fluid dynamics (CFD) code, CFX. The present study is carried out using CFD code to investigate cooling capability of a FM target cooled by natural convection. The steady state simulation as well as transient simulation is performed in the present work. Based on the transient simulation (T1), the minimum decay time that the maximum fuel temperature does not reach the design limit temperature (TONB-3 .deg. C) is around 15.60 seconds
[en] The Emergency Operating Guidelines (EOG) should be presented to provide technical information to prepare reactor-specific Emergency Operating Procedures (EOP) which cover operation during emergency events. Applicants for operating license and licensees of reactors under construction are required to: - Perform analyses of transients and accidents including multiple failures - Prepare emergency operating guidelines - Upgrade emergency procedures, including procedures for operating with natural circulation conditions - Conduct operator retraining The Procedure Generation Package (PGP) should be submitted to the regulatory body in Korea for the reactor licensing at the same time with Final Safety Analysis Report (FSAR), where PGP includes: - Reactor-Specific Technical Guidelines (RSTGs), - EOP Training Program - EOP Implementation Program The information should comply with those requirements associated with the development of EOG according to the Korean atomic law. All assumptions made in the EOP, which relate to safety analysis, must be verified to be true and appropriate for each user by each user. Furthermore, a set of EOP are required as one of operating procedures that shall be developed for all safety related operations that may be conducted over the lifetime of the facility in research reactors by international standards, Safety of Research Reactors (IAEA NS-R-4) : 7.51 (g) the reactor operator's response to anticipated operational occurrences and DBAs and, to the extent feasible, to BDBAs. In Korea, the regulation on codes and standards for nuclear facilities such as research reactors recommends the nuclear facilities operating organization to prepare EOP. Hence a set of EOG is proposed for helping to develop the EOP in a simplified manner for a research reactor. In this paper, it is described about a development and a revision of a set of EOG for a research reactor
[en] The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters
[en] Light water is the coolant that is stored in the reactor pool. For both waters, the cooling system for each is absolutely isolated and run independently. The classification is also different: safety system for cooling light water as coolant and non- nuclear safety for cooling heavy water as reflector. There will be, therefore, only negligible effect on the safety from any failures related to the cooling system for heavy water outside the reactor pool. Failures such as pipe ruptures in the heavy water system inside the reactor pool introduce, however, a different issue concerned with power control. Here the issues are presented and it will be shown that the safety of the research reactor is to be assured. A postulated initiating event related to the power control introduced by downgrading of the heavy water, i.e., a reflector was analyzed by simulating the reactor transient with the RELAP code. The result showed that the fuel integrity is assured using the reactor protection system with acceptable margin
[en] During startup operation, control absorber rods are not located in the equal critical position since they can be manually controlled by an operator without position limitation. Therefore, the power peaking factor in this control mode becomes larger due to the skewed power shape, making the accident consequence worse. In research reactor, the reactor protection system (RPS) has linear power trip and power lograte trip for a safe shutdown of reactor in the accident, and the occurrence of those trips depend both on the initial reactor power and the reactivity insertion rate. Therefore, with a series of sensitive analyses, we identified the most severe combination of initial conditions among the various initial reactor powers and reactivity insertion rates. The model reactor in this analysis is a 5MW pooltype research reactor having two different operation modes; a power operation and a training operation.. Since the accident occurs during startup of the reactor, the training mode without a forced convection results in more severe consequences in a view of fuel integrity. Therefore, the inadvertent withdrawal of a control rod during a startup of training operation is analyzed as a limiting case of the accident. Sensitivity tests with combinations of different initial reactor powers and reactivity insertion rates are performed for an inadvertent CAR withdrawal during startup of the training operation
[en] The procedures may also be used as an aid for assessing and documenting the results of tests. The commissioning procedures should include information that specifies several items. Those are mainly (1) all the activities and performance parameters that are to be measured under specified steady state and transient conditions, (2) the requirements on performance, together with clearly stated acceptance criteria. The final phase of stage C commissioning is reactor performance test, which is to prove the integrated performance (neutron power calibration, Control Absorber Rod drop time, I and C functioning, Rod worth, Core heat removal with natural mechanism) and the safety of the research reactor at full power with fuel loaded. Commissioning procedure was developed to show the safety of the research reactor. Both indirect and direct indicators were selected to show that the safety is ensured: 1) indirect parameters which imply success of safety functions: power, flow, opening valves, system response as-designed; 2) direct parameters which shows no failure of safety functions: no meaningful increase in level of neutron in the cooling system. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters. A preliminary comparison to conservative estimation has shown that the nuclear reactor safety of the research reactor will be assured by verifying that the reactor power and the PCS flow rate are conservative
[en] The fuel test loop of HANARO has been modeled with MARS code to predict the Peak Cladding Temperature (PCT) which is one of the design criteria for the design basis accidents. The PCTs have been calculated in the various operation parameters such as mass flow rate, coolant temperature, system pressure, the thermal power of test fuel, and the valves of safety function. In case of the room 1 LOCA the PCT increases with the increase of the thermal power, the coolant temperature, and the stroke time of the cold leg loop isolation valves as compared with that calculated in the design operation condition. However the effect of the stroke times of the safety injection valves and the depressurization valves on the PCT is negligible
[en] In research reactor, there is a reactor protection system (RPS) to keep the reactor in safe condition. In relation to the RIAs, the RPS has power trip and power lograte trip variables. The results of RIAs are affected by not only the initial reactor power but also the reactivity insertion rate. Therefore, we selected the parameters of the initial reactor power and the reactivity insertion rate for the sensitivity investigation in this study. Especially, a research reactor with a normal power of 5MW was considered. This sensitivity study was carried out by using RELAP5/MOD3.3. The CHFRs and fuel temperatures were investigated in the various initial reactor powers and reactivity insertions by using RELAP5/MOD3.3. In case of the step insertion of 1.5mk reactivity, the larger the initial reactor power is, the smaller the CHFR is. And in case of the constant reactivity insertion rates, there is a different trend with the initial reactor power. In case of the initial power of 100%FP, the minimum CHFR appears at the maximum reactivity insertion rate. On the other hand, in case of the initial power of 1.5%FP, the minimum CHFR is predicted at the minimum reactivity insertion rate
[en] The thermal hydraulic analysis of a research reactor building becomes much more important during longterm cooling stage in loss of normal electric power if the building is designed as containment to fulfill the enhanced regulation requirements of radiological consequence. Since the existing containment analysis computer codes are oriented to the condition and validated for nuclear power plants, it is necessary that a computer code adequate for the research reactors is developed. The purpose of this paper is to identify the thermal hydraulic phenomena during the long-term cooling of a research reactor and to select the appropriate analysis models and to use it as the basic data for the development of a computer code for reactor building and pool cooling analysis. This paper consists of three steps, and a description of each step is as follows: - PIRT for identification of thermal-hydrodynamic phenomena during long-term cooling of research reactors - Literature review and selection of analysis models for the thermal hydraulic phenomena determined by PIRT - Development of the code structure and algorithm of a computer code for cooling analysis of reactor buildings and pool.
[en] Highlights: • BEPU analysis were performed with a scenario of PCS pumps fail simultaneously. • The results from BEPU and conservative analysis were compared. • The comparing result shows the applicability and advantages of a BEPU safety analysis. - Abstract: Best estimate plus uncertainty (BEPU) is a promising approach to the safety analysis of nuclear reactors, and the uncertainty calculation is a very important concern about it. BEPU ensures realistic safety margins and secures higher reactor effectiveness by adopting best-estimate codes and realistic input data with uncertainties, whereas the previous conservative analysis generates excessive conservatism by considering each input parameter separately. A loss of flow accident (LOFA) of a 5 MW open-pool type research reactor was selected as a sample problem for a BEPU uncertainty assessment. We selected the failures of all primary cooling system (PCS) pumps, which would cause the abrupt reduction of flow and the reversal of core flow. The significant contributors to the reactor safety were identified and then input sets were sampled. For the uncertainty evaluation, 124 calculations were performed. This is the number of code runs required for a 95%/95% level with the 3rd order Wilk’s formula. The MOSAIQUE software developed by Korean Atomic Energy Research Institute (KAERI) was used for automated sampling of the uncertainty parameters, a global uncertainty calculation, and post processing of the results. The critical heat flux ratio (CHFR) and the fuel centerline temperature (FCT) were calculated at the 95%/95% level and were compared with those from conservative analyses. In addition, the impact of each design variables on the safety parameters was estimated by sensitivity analysis.