ULOF Accident Analysis for 300 MWt Pb-Bi Coolled MOX Fuelled SPINNOR Reactor

Transcription

1 ULOF Accident Analysis for 300 MWt Pb-Bi Coolled MOX Fuelled SPINNOR Reactor Ade afar Abdullah Electrical Enineerin Department, Faculty of Technoloy and Vocational Education Indonesia University of Education Jl. DR Setiabudhi no 229 Bandun Indonesia Tel Zaki Su ud Nuclear and Biophysic Research roup, Faculty of Mathematics and Natural Sciences Institut Teknoloi Bandun Jl. anesha 10 Bandun Indonesia Tel Rizal Kurniadi Nuclear and Biophysic Research roup, Faculty of Mathematics and Natural Sciences Institut Teknoloi Bandun Jl. anesha 10 Bandun Indonesia Tel Neny Kurniasih Nuclear and Biophysic Research roup, Faculty of Mathematics and Natural Sciences Institut Teknoloi Bandun Jl. anesha 10 Bandun Indonesia Tel Abstract In this study the result of investiation throuh simulation of unprotected loss of flow accident (ULOF) for 300 MWth MOX fuelled small Pb-Bi Cooled non-refuellin nuclear reactors (SPINNOR) are discussed. The two dimensional diffusion calculation combined with transient thermal hydraulic analysis has been employed. The reactor is tank type Pb-Bi cooled fast reactors with steam enerator included inside reactor vessel. The simulation bein with steady state calculation of neutron flux, power distribution and temperature distribution across the, hot and cool pool, and also steam enerator. The accident analysis bein with the loss of pumpin power. The sequence of analysis is then the transient flow calculation across the, the temperature distribution, hot pool and cool pool fluid dynamic calculation and temperature chane simulation, and the transient flow and temperature calculation across the steam enerator. Then the reactivity feedback calculation is conducted, followed by kinetic calculation, and then the sequence repeated. The results show that the SPINNOR reactor has inherent safety capability aainst this accident. Keywords: ULOF accident, Pb-Bi Cooled, SPINNOR reactor.

2 1. Introduction Since accident of Chernobyl, TMI II and the latest one Fukushima, the attention focuses on safety aspect of nuclear reactors. Therefore, development of safety analysis refers to inherent safety that claims simulation system which is more sophisticated. Wherever practicable, inherent safety characteristics have been incorporated into the desin of systems important to safety, particularly that fulfill the three basic safety functions of reactor shutdown, decay heat removal and the containment of radioactive materials (Zaki at al.,1996). The SPINNOR (Small Power Reactor, Indonesia, No On-Site Refuelin) are concepts of small lead-bismuth cooled nuclear power reactors with fast neutron spectrum that could be operated for more than 15 years without on-site refuelin. They are based on the concept of a lon-life reactor developed in Indonesia since early 1990 in collaboration with the Research Laboratory for Nuclear Reactors of The Tokyo Institute of Technoloy (Zaki, 2007). The reactor is used in relatively isolated areas, preferably Small Islands, and operated to the end of its life without refuelin or fuel shufflin. Some important requisite characteristics are easy operation, easy maintenance, transportability, inherent/passive safety and nuclear proliferation resistance. In this paper we analize the safety performance of SPINNOR reactor. By the term safe here we mean that reactor is able to survive ULOF (Unprotected Loss of Flow) accident without reactor scram or the help of the operator. 2. Desin Concept of SPINNOR Reactor A schematic view of the system bein analyzed is shown in Fi.1. The overall system includes a reactor, hot pool, cool pool, steam enerator and pump. The intermediate heat exchaner (IHX) is eliminated, and heat from primary coolant system is transferred directly to the steam-water loop throuh the steam enerator. The coolant flows trouh the removes heat enerated in the, and then flows up to the hot pool. From the hot pool, coolant flows into the steam enerator, transferrin the heat into the steam-water loop, and then oes down to the cool pool. From the cool pool the coolant is pumped back to the (Zaki, 2007). 3. Calculation Model 3.1. Calculation of Neutron Distribution Flux If we evaluate reactor in the steady state, then time variable can be eliminated. We assume homoeneous material reion and adopt diffusion equation in this model (Duderstadt at al.,1976):, ( ) ( ) x D r r ( r ) ( r ) r f 1 s k 1 1 eff (1) D r keff s = Diffusion constant for roup = Neutron flux for roup = Cross-section removal for roup = Fission spectrum fraction = Effective critically factor = Number of neutron produced per fission = Macroscopic cross-section scatterin from roup to roup After some simplification one of reasonable model for the fast reactor safety analysis is the adiabatic mode with the shape function becomes : x, D ( r, ( r, ( r, ( r, (2) 9 9 r f 1 s k 1 1 eff Diffusion equation of multi roup is solved in numerical with method SOR (Succesive Over Relaxation). Neutron population in reactor durin transient process is determined by solvin of point kinetics equation. If transformation of spatial distribution can be eliminated, then we can obtain level of reactor power as function time p( by solvin of point kinetics equation as follows, (Duderstadt at al.,1976):

3 C i = Reactivity (dk/k) = Effective delay neutron fraction = Delayed neutron fraction = Delay neutron precursor concentration = Neutron eneration time 3.2. Thermal Hydraulic Calculation 6 dp( ( p( ic i dt i1 (3) dci ( i p( ici ( dt (4) Calculation of thermal hydraulic covers calculation distribution of temperature in all parts of reactor that is in reactor and in steam enerator ( S) that is fuel temperature, coolant temperature and temperature claddin and ap, so pressure in reactor and all coolant circulation aspects in reactor. Fiure 2 shows thermal hydraulic model in this calculations. Reactor is divided into concentric rin, where the cross-flow between two adjacent rins is assumed zero. For coolant, calculation used mass and enery conservation equation: w c p = Mass density = coolant mass flow rate = Specific heat at constant pressure Q = Power density P f D e T t T z ''' c p wc p Q (5) Pressure drop calculation is solved momentum conservation equation. = Total mass flow = Pressure = friction factor = Effective hydraulic diameter = ravitational acceleration P f t z z 2D 2 e 2 (6) 3.3. Hot pool and Cool Pool Durin this analysis, averae temperature in both hot pool and cool pool are used. old T [( h A C T tc T ]/[( h A t C ] (7) ps p p ps p h i1 ( h j A ( ps ) /( A ) (8) T h A ps C p = Temperature of hot pool = Heiht of hot pool = Area of hot pool = Mass density of hot pool = Total mass flow rate in primary S = Specific heat of hot pool = Total mass flow rate of

4 C p = Specific heat of Similarly with averae hot pool, temperature cool pool is solved these equations: old T [( h A C T tc T ] /[( h A t ps p p ps h i1 C ] ( h j A p ( ps ) /( A ) (9) (10) T = Temperature of cool pool h = Heiht of cool pool A = Area of cool pool = Mass density of cool pool 3.4. Feedback Calculations ( in Eq. (3) is summed external reactivity and feedback reactivity. Feedback reactivity is included Doppler, axial expansion, radial expansion, and void reactivity. Feedback reactivity is the effect of the temperature T. We can express temperature coefficient of reactivity: (11) T is a neative value that we inserted in calculation since increasin temperature will cause decreasin. T is averae temperature in (fuel and coolan eneral Calculation Alorithm Calculation method employed in the present study is two dimensional r-z eometry for diffusion calculation is carried out every year to et neutron flux distribution and power distribution. The flow diaram of the eneral calculation alorithm is iven in Fi. 3. For the first step, the steady state neutronic and the thermal-hydraulic calculation are performed, and then the accident condition is taken. The ULOF accident is started by failure of the primary pumpin system. Temperature in will be chane depend on how much flow rate has inserted. Chanin of temperature will chane averae hot pool temperature and steam enerator temperature. In this research reactivity external of reactor is assumed constant in every time. Heat will be transferred to water circulation in steam enerator. 4. Result and Discussion Table 1 shows the main reactor desin parameter. Fi. 4 shows the reactor power chane pattern durin accident. The power increases at the beinnin of the accident, then radually decreases and stabilizes around 40% of initial power. Fi.5 shows temperature of fuel, claddin and coolant. The maximum fuel temperature was about 800,23 o C, the maximum claddin temperature about 634,07 o C, and the maximum coolant temperature about 627,89 o C. Fuel, claddin and coolant temperature are still below the maximum temperature limits. Fi. 7 Show flow rate chane durin ULOF accident, there are also small oscillation in total flow and primary S flow rate, which are mainly caused by the fact that flow rate in the and primary side S in the absence of primary pumpin is mainly controlled by level and temperature of coolant is the hot and cool pool which in turn depend on the flow rate of the and primary S. 5. Conclusions The conclusions are summarized as follow, 1. In eneral the reactor can be survived the ULOF accident inherently. 2. The relatively hih natural circulation component contribute important factor to this ability to survive ULOF accident. 3. Accident analysis result show of the case maximum coolant, fuel and claddin temperature are still below each temperature limit, and marin to coolant and fuel are lare.

5 References Zaki Su ud (2007), Advanced SPINNORs Concept and The Prospect of Their Deployment in Remote Area, International Conference on Advances in Nuclear Science and Enineerin, Bandun, Indonesia, pp S. Zaki and H. Sekimoto (1995), Safety Aspect of Lon-Life Small Safe Power Reactors, Ann. Nucl. Enery, Vol. 22, No 11,pp Zaki Su ud (1998), Comparative Study on Safety Performance of Nitride Fueled Lead Bismuth Cooled Fast Reactor With Various Power Levels, Proress in Nuclear Enery, Vol. 32, No. 34, pp S. Zaki and H. Sekimoto (1996), Accident Analysis of Lead-Bismuth Cooled Small Safe Lon-Life Fast Reactor Usin Metallic or Nitride Fuel, Nuclear Enineerin and Desin 162, pp Waltar A. E. and Reynolds A. B (1981), Duderstadt J.J. and Hamilton L. J. (1976), Fast Breeder Reactor, Peramon Press. Nuclear Reactor Analysis, John Willey and Sons. Table 1. Main reactor desin parameter No Parameter Value 1 Reactor Power 300 MWth 2 Fuel UO 2 - PUO Shieldin Material Coolant Pin/Pitch diameter Pin diameter Claddin Thickness Total Flow rate B4C Pb-Bi 1.2 cm 1.0 cm 0.8 cm k/s steam Reactor Vessel Subcooled water Hot pool Pump shieldin reflector Steam enerator (S) inlet Fiure 1. Overview of the SPINNOR reactor

6 CENTER OF PELLET FUEL-CLADDIN AP (COOLANT) FUEL PELLET CLADDIN WIRE SPACER CENTER OF COOLANT CHANNEL COOLANT SPATIAL MESH FOR THERMAL HYDRAULIC CALCULATION Fiure 2. Calculation model for coolant channel Fiure 3. Calculational flow diaram Fiure 4. Power chane durin ULOF accident.

7 Fiure 5. Hot spot temperature durin ULOF accident Fiure 6. Reactivity component chane durin ULOF accident Fiure 7. Flow rate chane durin ULOF accident