Cognitive Approach to Severe Accident in Nuclear Power Plant Using MAAP4

Transcription

1 Transactions of the 17 th International Conference on Structural Mechanics in Reactor Technology (SMiRT 17) Prague, Czech Republic, August 17 22, 200 Paper # WP01- Cognitive Approach to Severe Accident in Nuclear Power Plant Using MAAP4 Seung Dong Lee, Kune Y. Suh 1), Goon Cherl Park, Un Chul Lee 1) Department of Nuclear Engineering, Seoul National University ABSTRACT Use of instruments in severe accident guidance is focused on cooling of the reactor core and heat removal from the containment. The large uncertainties and extreme conditions involving the severe accident render the measurement and interpretation of the data difficult enough. There presently exist some parameters relevant to monitoring the accident initiation and progression and to preparing necessary measures and strategies to prevent further aggravation and to mitigate the consequence of the accident at hand. One of the accident management guidance parameters for the KSNP (Korean Standard Nuclear Plant) is the core exit temperature (CET). Though the fuel rod temperature is perhaps the most important parameter to determine the phase of accident progression, there is currently no means to read it directly from the core. A best alternative, thus, is to guess the fuel temperature from the measured CET data. The severe accident management guidance (SAMG) suggest that the transition criterion from the emergency operating procedure is the CET of 60 K. The fuel temperature begins to escalate from the time of core uncovery. When the temperature reaches 90 K, hydrogen is generated by oxidation of the cladding whose reaction is accelerated drastically with temperatures exceeding 100 K. Because the oxidation reaction is exothermal, the core temperature increases rapidly. Although the CET is lower than the fuel temperature, the increase rate will tend to follow that for the fuel rod temperature. In this paper, the focus is placed on the plant damage state during a severe accident. The accident initiator is a small-break loss-of-coolant accident (LOCA). The MAAP4 calculations were carried out to analyze the plant state and accident sequence such as the core uncovery time, the CET, and the primary system pressure. It is concluded that the CET will prove to be a reasonable criterion for recognizing severe core damage. However, when the temperature exceeds 90 K the CET reading may not be reliable so that one needs alternatively to infer the degree of core damage by reading the reactor water level and the hydrogen generation rate. KEY WORDS: severe accident, core exit temperature (CET), loss-of-coolant accident (LOCA), core uncovery, core damage, hydrogen generation, severe accident management guidance (SAMG), accident mitigation strategy INTRODUCTION Plant instruments play an important role in developing severe accident management guidance (SAMG) [1-]. Although the instruments are designed essentially for the design basis accident, they may give useful information on which strategy is needed and if the strategy may be successful in a severe accident environment and the operator can identify safety challenges by them. Exact analysis of the instruments is based on the accident mitigation strategies. Because severe accidents involve a variety of scenarios and a great deal of uncertainties, the instruments also carry much uncertainty in terms of their availability. Use of instruments in the SAMG is focused on cooling of the reactor core and heat removal from the containment. Fuel temperature is one of the most important parameters, from which one may surmise the degree of accident progression. But there is no way to measure the fuel temperature directly. One can thus infer the fuel temperature by detecting the core exit temperature (CET). The SAMG for the Korean Standard Nuclear Power Plant [4,] presents that the reference point is the CET of over 60 K. The fuel temperature rises from core uncovery and, when it reaches 90 K, hydrogen generation begins by oxidation. Because the fuel cladding oxidation involves an exothermic reaction, the fuel temperature rises drastically over 10 K. The objective of the emergency operating procedure (EOP) is to cool the core, whereas that of the SAMG is to prevent fission product release out of the containment. In order to continually deal with accident progression in the EOP phase a transition to SAMG is extended. RUN CASES The reference plant as shown in Figure 1 is the Ulchin Units &4 [6], which has the total free volume of m, the inside diameter of 4.9m, the design internal pressure of 7psig and the nominal power of 281MWth. The plant has a large dry containment, two steam generators and four cold legs. Figure 2 presents that the containment is nodalized into five () subcompartments such as cavity, lower, upper, annular and dome compartments. In this nodalization six junctions and one failure junction are used. There are some important accident initiators, which contribute to core damage frequency as listed in Table 1. The small-break loss-of-coolant accident (LOCA), which is one of the dominant accident sequences in the reference plant, is selected as the example accident scenario. The break is located 7.71m from the reactor vessel bottom in the cold leg. Major engineered safety features are supposed to have 1

2 failed including the high pressure safety injection system, low pressure safety injection system, and containment spray system. Thirteen cases were analyzed with break sizes varying from m 2 to m 2. The run time for the calculation was 24hours each. Table 1. Core damage frequency contributions (Taken from Ref. 6) Initiating Events Core Damage Frequency Mean ( /RY) Error factor % of total 1. Large-Break LOCA 1.0E Medium-Break LOCA 6.E Small-Break LOCA 1.86E Steam Generator Tube Rupture 1.14E Vessel Rupture 2.66E Interfacing System LOCA 1.77E General Transients.9E Loss of Feedwater 1.14E Loss of Condenser Vacuum 2.E Loss of a 4.16kV AC Bus.48E Loss of a 12V DC Bus.17E Loss of Off-site Power.97E Station Blackout 4.80E Large Secondary Side Break 1.46E Anticipated Transient without Scram.1E LOCCW 1.2E Total 8.2E Dome 6 EL. 20' 7 Upper EL. 167' 4 EL. 167' Annular Lower 1 Lower Annular 2 1 Cavity EL. 86' EL. ' Fig. 1 Side view of reference plant RESULTS AND DISCUSSION Fig. 2 Containment nodalization The parameters concerned with the core damage indicators were divided into two groups. One is the in-vessel group such as the reactor water level and the primary system pressure, while the other includes the ex-vessel group such as the containment pressure and temperature. Figure demonstrates that the reactor water level decrease rate increases linearly with the break size. Figure 4 illustrates that the depressurization rate increases with the break size. If one has 2

3 the knowledge of the water level decrease rate and the system depressurization rate simultaneously, one can infer the core temperature increase rate and finally the core melting time utilizing both the graphs and the CET. Figure shows two maximum core temperature trends. One is the temperature at core uncovery, and the other is the temperature at the reference point where the CET reaches 60 K. The temperature at core uncovery indicates an increasing trend and the temperature at the reference point indicates a decreasing trend. This phenomenon may well be explained as follows. Although the core temperature rise rate has an increasing tendency as shown in Figure 6, it takes less time from core uncovery to the reference point as the break size increases. Hence the core heatup time is extended and the maximum core temperature is lower at the reference point as the break size increases. In this work the containment temperature is chosen as the ex-vessel core damage indicator. Figure 7 also indicates the increasing trend but the temperature rise rate is mild so that the operator may not have to identify the core damage state immediately. In this study the core uncovery time is defined as t uncovery, the reference CET as t CET, and the cladding rupture time as t clad. Observe from Figure 8 that the time duration from core uncovery to the reference point is longer than that from the reference point to the cladding rupture time. But the difference shrinks as the break size increases and from about m 2 the time duration is in fact reversed Water level decrease rate (m/sec) Water level decrease rate Fig. Water level decrease rate from core uncovery to reference point 1600 Depressurization rate (Pa/sec) Depressurization rate -200 Fig. 4 Depressurization rate from core uncovery to reference point

4 Maximum core temperature (K) Maximum core temperature at core uncovery time Maximum core temperature at reference point Maximum core temperature (K) Fig. Maximum core temperature 0.40 Core temperature rise rate (K/sec) Fig. 6 Core temperature rise rate from core uncovery to reference point Containment temperature rise rate (K/sec) Fig. 7 Containment temperature rise rate from core uncovery to reference point 4

5 t CET -t uncovery t clad -t CET Time (sec) Fig. 8 Effect of break size on time difference CONCLUSION Several physical parameters were examined to use in prospective application to SAMG for the KSNP. The smallbreak LOCA scenario was selected as the accident initiator and such plant status was studied as the system pressure, the core uncovery time and the core melting time. Although the core damage indicators were divided into the in-vessel and ex-vessel groups, the current work was focused on the CET, because the operator may not have to identify the core damage state immediately from other parameters such as the reactor water level, the primary system pressure and the containment pressure and temperature. Based on the MAAP4 calculation the time duration was examined from the core uncovery to the SAMG point and also from the reference point to the core melting time. As the break size increases, it takes less time from the SAMG point to the core melt so that the operators must be alert to more rapidly execute necessary accident mitigation strategies. REFERENCES 1. Hanson D. J., et al., Accident Management Information Needs: Methodology Development and Application to a Pressurized Water Reactor with a Large, Dry Containment, NUREG/CR-1, Vol. 1&2, Idaho Falls, ID, USA, April Severe Accident Management Guidance, Westinghouse Electric Corp., Pittsburgh, PA, USA, Vols. 1&, June Hanson D. J., et al., Assessing Information Needs and Instrument Availability for a Pressurized Water Reactor during Severe Accidents, Nuclear Engineering and Design, Vol. 148, 1994, pp Develoment of Accident Management Technology and Computer Codes, Korea Atomic Energy Research Institute, Taejeon, Korea, Development of Accident Management Guidance for Korean Standard Nuclear Power Plant, Korea Atomic Energy Research Institute, Taejeon, Korea, Ulchin Units &4 Final Probabilistic Safety Assessment Report, Korea Electric Power Corp., Seoul, Korea, February 1994.