3. Boiling Water Reactors - Chapter 3 BOILING WATER...

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Unformatted text preview: Chapter 3 BOILING WATER REACTORS © M. Ragheb 11/2/2008 3.1 INTRODUCTION Boiling Water Reactors (BWRs) operate at a reactor vessel pressure of 1,040 psia, a value considerably lower at almost one half that the operating pressure of PWRs at 2250 psia. Their nuclear steam supply system is different than PWRs in that the steam is produced within the core and is directly fed to the turbine-generator plant. This system uses a direct cycle as opposed to the indirect cycles used in other nuclear power plants. BWRs are the most commonly deployed design after the PWR design. The Clinton BWR plant using an artificial cooling late in Central Illinois, USA is shown in Fig. 1. Fig. 1: The Clinton Boiling Water Reactor with its cooling lake in Central Illinois, USA. 3.2 BOILING WATER REACTOR POWER CYCLE Boiling Water reactors with a direct cycle offer the capital cost advantage of eliminating the need for steam generators and a pressurizer. Another feature is the heat transfer in the core is mostly by latent heat as opposed to sensible heat in other power plants. This results in smaller flow rates and pumping energy needs. These advantages are countered by the presence of the short-lived isotope N16 in the turbine plant area. Nitrogen16 has a half-life of 7.1 seconds with beta emissions at energies of 4.3 and 10.42 MeV, and Gamma emissions at 6.129 and 7.115 MeV. This prevents access to parts of the turbine area for maintenance until the system is shutdown to allow for N16 decay. Additional shielding for the piping, turbine and feed water heaters is needed. Design features which provide a long transit time for the steam to move in the main steam pipe to the top of the reactor building then to the bottom, before getting into the turbine building alleviate the problem. Skyshine of the gamma radiation on the outside of the turbine hall could still be present during operation. Fig. 2: Power cycle of the BWR-5 plant. Fig. 3: Light Bulb BWR containment structure showing the pressure suppression pool. The coolant passing the reactor core requires strict water chemistry management to avoid the activation of the carried corrosion products. The activated corrosion products would otherwise accumulate in piping hot spot, increasing the radiation dose to the operating personnel. The coolant passing through the core finds 10 percent of it converted into steam. As shown in Fig. 2, the wet steam goes through a set of steam dryers and separators above the core, which return the separated water to the core. Circulation pumps, as shown in Fig. 3 circulate the coolant back to the core. The steam is passed from the top of the core to the steam turbine. Because of the presence of the steam drying equipment at the top of the reactor vessel, the control rods in the BWR (Fig. 4) are inserted from the bottom of the core, in contrast to the PWR system where they are inserted from the top. They contain cooled tubes with boron carbide (B4C) as a neutron absorber. The control rods are hydraulically driven by the reactor coolant pressure. Another reason for the control rods to be inserted from the bottom is that they possess there a higher worth since the void fraction is lower near the bottom of the core resulting in higher neutron moderation, higher neutron flux, and higher power level. One would like to control the power level where it peaks closer to the bottom of the core (Fig. 5). Fig. 4: BWR control rod drive. Fig. 5: BWR cutout view showing core, steam dryers and control rod assembly. 3.3 CONTAINMENT STRUCTURE The BWR vessel is surrounded in a containment structure equipped with a pressure suppression pool shown in Figs. 3 for a light-bulb containment design, and in Fig. 6 for a steel shell and concrete containment design. The positioning of the pressure suppression pool below the reactor core does not allow for natural circulation convective cooling in the case of a loss of coolant accident (LOCA). More advanced inherently safe designs position the pressure suppression pool above the core. In case of an accident, the pressure in the core and the pressure suppression pool are equalized by the operator allowing natural circulation from the core to the pressure suppression pool in this case. Fig. 6: Steel shell and concrete BWR containment structure showing the pressure suppression pool. 3.4 CORE VESSEL The reactor vessel for the BWR/6 is shown in Fig. 3. Table 1 gives the operational characteristics of this design. Table 1: Operational Characteristics of a typical BWR. Characteristic Value Thermal Power output System pressure Fuel enrichment Coolant flow Core inlet enthalpy Average exit quality Core average void fraction Maximum exit void fraction Average inlet velocity Core pressure drop Inlet temperature, feed water Core inlet temperature Outlet temperature Maximum fuel temperature Average linear heat rate Maximum linear heat rate Average heat flux Maximum heat flux Minimum CHFR Active height Equivalent active diameter Height to diameter ratio Active core volume Average core power density Fuel weight Specific power Burnup Conversion ratio Number of fuel assemblies Fuel element array Assembly dimensions Assembly pitch Number of fuel rods per assembly Total number of fuel rods Fuel rod outer diameter Fuel rod pitch Pitch to diameter ratio Cladding thickness Fuel pellet diameter Pellet to clad gap size Pellet density, percent of theoretical Fission gas plenum length 3.5 FUEL ASSEMBLIES 3,579 MWth 1,040 psia 2.2-2.7 % 1.05x108 lbs/hr 527.9 Btu/lb 14.6 % 42.6 % 76 % 7.1 ft/sec 25.9 psi 420 oF 532 oF 547 oF 3,500 oF 6.0 kW/ft 13.4 kW/ft 159,000 BTU/(hr.ft2) 354,000 BTU/(hr.ft2) >1.9 148 inches 144 inches 1.03 2260 ft3 1580 kW/ft3 138,000 kgs 25.9 kW/kg U 27,500 MW.days/MTU 0.5 732 8x8 5.52 in x 5.52 in 6.0 in 62 45,384 0.493 in 0.64 in 1.3 0.034 in 0.416 in 0.0045 in 94 % 12 in The nuclear fuel assemblies are positioned inside a core shroud within the reactor vessel. Subcooled water is introduced at the bottom of the core and flows upward through vertical fuel assemblies (Fig. 7). The fuel rods are arranged in core with a cross sectional area equivalent to a circle of 12 feet diameter. The core consists of 700 fuel assemblies with a square cross section 5.52 inches on the side. Groups of four fuel assemblies are arranged around the cruciform control rod. Fuel rods are arranged in an 8x8 array. They contain 62 active fuel rods and 2 hollow watyer rods which are supported by the upper and lower tie plates. Each fuel assembly is surrounded by a Zircaloy-4 channel. The presence of this channel prevents the flow between adjacent fuel assemblies. The flow to individual fuel assemblies can thus be orificed so as to maintain a uniform exit quality from all fuel assemblies. Fig. 7: BWR fuel assembly. The fuel consists of UO2 pellets with an enrichment varying from 2.2 to 2.7 percent in U235. The fuel pellets are 0.416 in in diameter and 0.41 in in length. The pellets are contained in in Zircaloy-2 tube with a wall thickness of 0,p034 in. The fuel pellets are at a clearance from the cladding of 0.0045 in. A two phase mixture at about 14 percent quality exits from the fuel assemblies and enters the upper plenum region in a domed top. An array of standpipes is attached to the dome with a three stage steam saparator at the top of each standpipe. The centrifugal force separates the water from the mixture in each stage. 3.6 JET PUMPS The water in the pool surrounding the standpipes is mixed with feed water by means of the feed water sparger. The water then flows to the down-comer annulus where the coolant circulation inside the core is achieved by the use of a set of 20 jet pumps. Fig. 8: BWR Pressure vessel and jet pumps. The jet pumps (Fig. 9) are mounted internally inside the reactor core and have no moving parts inside the reactor vessel. Water flow is induced through the Bernoulli effect generating a suction pressure inside the jet pump nozzle with a secondary flow by the coolant pumps pumps positioned outside the reactor vessel (Fig. 10). Fig. 9: BWR jet pumps. Fig. 10: Bernoulli effect generates suction in BWR jet pump, without moving parts. 3.7 FLOW SYSTEM The coolant recirculation system consists of two loops external to the reactor vessel. One third of the recirculation water enters the recirculation pumps. The recirculation water prfessure is increased by the pumps and is distributed to a manifold from which connecting pipes are welded to the recirculation water inlet nozzles of the rector vessel. One inlet nozzle serves two jet pumps Since the BWR design does not have a separate stem generator like the PWR design, water chemistry becomes paramount to avoid corrosion and activation of the corrosion products. These activated products can accumulate at hot spots in the reactor circuitry leading to a radiation dose to the operating personnel. The flow diagram for the Quad Cities BWR is shown in Fig. 11. It shows the cleanup demineralizer system. A moisture separator is used between the high pressure and low pressure turbines. The condensate storage tank is used to operate the control rod drives and serves as a supply of cooling water in the case of an emergency. Fig. 11: Quad Cities BWR flow diagram. 3.8 BOILING TUBE REACTOR DESIGN The boiling tube reactor design is used in the RBMK-1000 Russian design (Fig. **). It uses light water as the boiling medium and graphite as a moderator. The steam is separated from the two phase mixture in steam drums. The multiple tubes design offers a level of safety since in the case of coolant leakage in a tube it could be isolated from the rest of the system. However the use of light water as a coolant, and its acting as a neutron poison leads to a positive power coefficient of reactivity. When the temperature increases leading to water boiling, this corresponds to a positive reactivity insertion, and consequent power level rise. This positive feedeback effect, can make the system unstable at low power level. This type of behavior contributed to the Chernobyl accident which occurred in this type of design. Fig. 12: RBMK-1000 Boiling light water graphite moderated tube reactor. Steam is separated from two phase mixture in steam drums. Fig. 13: Top view of RBMK-1000 core during on line refueling. 3.9 ADVANCED BWR The Advance Boiling Water Reactor (ABWR) is an evolutionary design providing a higher chimney and consequently better natural convection cooling in the core (Fig. 14). The Advanced Boiling Water Reactor (ABWR) already has track record. In Japan, four ABWR units are in operation; another three units are under construction in Taiwan and Japan, and nine more units are planned in Japan. It is feasible that an ABWR plant could be built in the USA and be commercially operational by 2012. The design is available today for immediate generation needs of about 1500 MWe, providing technology and schedule certainty, along with competitive economics. The ABWR is a direct cycle BWR that reflects 50 years of continued evolution from GE's initial BWR concept. It combines the best features of GE's worldwide BWR fleet with advanced technology enhancements, such as digital controls, that improve performance and longevity. The ABWR design is already licensed in three countries: the USA, Japan and Taiwan. Fig. 14: ABWR Plant Design. Source: GE. Fig. 15: Pressure vessel of the Advanced Boiling Water Reactor, ABWR. 3.10 ECONOMICAL SMALL BOILING WATER REACTOR, ESBWR The next evolution of advanced BWR technology is the ESBWR. It utilizes a number of new features to provide better plant security; improved safety; more location options; excellent economics; and operational flexibility that ultimately increases plant availability. Fig. 16: ESBWR pressure vessel. Source: GE. Some of its primary features are: 1. Simplified design features: Passively removes decay heat directly to the atmosphere. Eleven systems are eliminated from the previous designs. It involves the use of 25 percent fewer pumps, valves and motors than other BWR designs. 2. Passive design features reduce the number of active systems, increasing safety: It is 11 times more likely for the largest asteroid near the earth to impact the earth over the next 100 years than for an ESBWR operational event to result in the release of fission products to the environment. 3. Incorporation of features used in other operationally-proven reactors, including passive containment cooling, isolation condensers, natural circulation and debris resistant fuel. 4. Natural circulation and convection cooling is used. 5. Expedited construction schedule because of standardized modules and pre-licensed design. Estimated construction times for first concrete to first core load of 36 months 3.11 COOLANT ACTIVATION AND N16 FORMATION INTRODUCTION In light water and heavy water moderated and cooled reactors, the threshold fast neutron activation set of reactions with the isotopes of oxygen in the water: 8 O16 + 0 n1 → 1 H 1 + 7 N 16 16 → 8 O16 + −1 e0 + γ 7N with: T1 2 ( 7 (1) N 16 ) = 7.1sec , Eγ = 6.13 MeV in 69 percent of the decays, σactivation = 46 mb, abundance of 8O16 = 99.758 percent, ρ ( H 2O) = 1.0 [ gm / cm3 ] and the set of reactions: 8 7 T1 2 ( 7 N 17 → 8 O17* + −1 e0 + γ 8 with: O17 + 0 n1 → 1 H 1 + 7 N 17 O17* → 8 O16 + 0 n1 (2) N 17 ) = 4.17sec , σactivation = 5.9 mb, abundance of 8O17 = 0.038 percent, * denotes an excited state, are significant, particularly in the Boiling Water Reactor (BWR), because of the short transit time of the generated steam between the reactor core and the turbine and other equipment external to the reactor shield. RATE EQUATIONS The rate equations describing the production and decay of N16 and N17 are respectively: 16 dN 16 = R( 8 O16 ) − λ N N 16 dt 17 dN 17 = R( 8 O17 ) − λ N N 17 dt (3) dN i = Ri − λi N i , i = 1, 2. dt (4) or: SOLUTION OF THE RATE EQUATIONS FOR ASINGLE LOOP THROUGH THE REACTOR For a single loop through the reactor, we solve the rate equations accounting for the production and decay rates of the 7N16 and 7N17 in nuclei per unit volume, assuming their initial concentrations at time t=0 to be negligible. By multiplying both sides of Eqn. 4 by an integrating factor: eλit we get: dN i = eλit Ri − eλit λi N i dt dN i + λi N i .eλi t = Ri .eλit eλit dt eλit The left hand side can be written as a total differential as: d ( N i .eλit ) = Ri .eλit dt (5) which can be proved by applying the chain rule of differentiation. Separating the variable and integrating using limit integration, we get: Ni ( t ) ∫ t d ( N i .eλi t ) = Ri .∫ eλit dt Ni 0 N i (t )eλit − N io .1 = Multiplying both sides by: e − λit 0 Ri λi ( eλit − 1) N i (t ) = N io .e − λit + Ri λi (1 − e − λit ) (6) For negligible nuclei at time t = 0, we get: Ri N i (t ) = λi (1 − e − λit ) (7) The activities of the species can be written as: Ai (t ) = λi N i (t ) = Ri (1 − e − λit ), i = 1, 2 (8) Writing Eqn. 8 explicitly yields for the activities after a single loop through the reactor: A16 (t ) = λ16 N N 16 (t ) = R16 (1 − e − λ16t ) A17 (t ) = λ17 N N 17 (t ) = R17 (1 − e − λ16t ) (9) PRODUCTION RATES The rates of production of N16 and N17 per unit volume can be calculated from: Ri = Σ activation ,iφ = N iσ activation ,iφ n cm .sec is the microscopic activation cross section [b] φ is the average neutron flux [ where: σ activation ,i (10) 2 Σ activation ,i is the macroscopic activation cross section [cm-1 ] Using the modified form of Avogadro’s law: N i = ai ρ Av M (11) ai are the percentage natural abundances A v is Avogadro's number where: gm cm3 M is the molecular weight of water [amu] ρ is the density of water [ Substituting Eqn. 11 into Eqn. 10 yields: Ri = ai ρ Av M σ activation ,iφ For the data of N16 and N17 and a neutron flux of φ = 1010 [ (12) n: cm .sec 2 99.758 1x0.6 x1024 nuclei 46 x10−3 x10−24 x1010 = 1.53x107 [ 3 100 18 cm .sec 0.038 1x0.6 x1024 nuclei R17 = 5.9 x10−3 x10−24 x1010 = 7.47 x102 [ 3 100 18 cm .sec R16 = It can be readily noticed that the production rate of N16 exceeds that of N17: R16 > R17 R16 1.53x107 = = 2.05 x104 2 R17 7.47 x10 ONE PATH THROUGH CORE AND OUTSIDE LOOP If the transit time through the reactor is tc, the activity after one loop in the reactor loop would be from Eqn. 8: Ai (t ) = λi N i (t ) = Ri (1 − e − λitc ) (13) After one loop in the reactor core and a time to in the outside loop, the total transit time is: T = t c + t0 (14) The activity acquired after one full path through the reactor core and outside loop, where it decays is becomes: Ai1 = Ri (1 − e − λitc )e − λito (15) MULTIPLE PATHS ACTIVATION After two paths through the core and the outside loop the activities would be from Eqn. 15: Ai 2 = Ri (1 − e − λitc )e − λi t0 e − λiT + Ri (1 − e − λi tc )e − λit0 = Ri (1 − e − λitc )e − λi ( to +T ) + Ri (1 − e − λi tc )e − λit0 (16) = Ri (1 − e − λitc )e − λit0 (1 + e − λiT ) After n loops Eqn. 16 can be generalized as: Ain (t ) = Ri (1 − e − λitc )e − λi t0 (1 + e − λiT + ... + e − ( n −1) λiT ) (17) Since: n ∑ e− ix = i =0 1 − e − nx = (1 + e − x + ... + e − nx ) −x 1− e (18) With x ≡ λiT it becomes: n −1 ∑ e− iλiT = i =0 1 − e − ( n −1) λiT = (1 + e − λiT + ... + e − ( n −1) λiT ) − λiT 1− e (18)’ Then we can express Eqn. 17 for the activation after n loops as: Ain = Ri (1 − e − λitc )e − λit0 1 − e − ( n −1) λiT 1 − e − λiT EQUILIBRIUM ACTIVITIES If the half-lives of the nuclides are not very long: (n-1)λiT = (n-1) ln 2 T >> 0 , T1 ,i 2 and consequently as n → ∞ , Eqn. 19 reduces to: (19) Ai∞ ≈ Ri (1 − e − λitc )e − λit0 1− 0 1 − e − λiT (20) 1 − e − λitc − λit0 e ≈ Ri 1 − e − λiT The equilibrium activities for N16 and N17 for: tc ≈ 2sec, to ≈ 5sec, T ≈ tc + to = 2 + 5 = 7sec. become: A16 ∞ ≈ 1.53x10 7 1− e 1− e A17 ∞ ≈ 7.47 x10 − ln 2 2 7.1 ln 2 7 − 7.1 2 1− e 1− e e − ln 2 2 4.17 − e ln 2 7 − 4.17 ln 2 5 7.1 − = 1.53x107 ln 2 5 4.17 Bq 0.177 0.614 = 3.360 x106 [ 3 ] cm 0.495 = 7.47 x102 Bq 0.283 0.436 = 1.340 x102 [ 3 ] cm 0.688 DECREASING THE EQUILIBRIUM ACTIVITIES A possible design feature in Boiling Water Reactors (BWRs) is to divert the steam from the reactor to the top of the reactor building then to its bottom before feeding it to the turbine. This would increase the outside loop transit time to to = 10 seconds: tc ≈ 2sec, to ≈ 10sec, T ≈ tc + to = 2 + 10 = 12sec. In this case the equilibrium activities would become: A16 ∞ ≈ 1.53x107 1− e 1− e A17 ∞ ≈ 7.47 x10 2 − 1− e 1− e ln 2 2 7.1 − ln 2 12 7.1 − e − ln 2 2 4.17 ln 2 12 − 4.17 e ln 2 10 7.1 − = 1.53x107 ln 2 10 4.17 Bq 0.177 0.377 = 1.480 x106 [ 3 ] cm 0.690 = 7.47 x102 Bq 0.283 0.190 = 4.649 x101[ 3 ] cm 0.864 Thus doubling the outside loop transit time from 5 to 10 seconds, reduces the equilibrium activities for the two nitrogen isotopes by factors of: 3.360 x106 = 2.3 1.480 x106 1.340 x102 N 17 : = 2.9 4.649 x101 N 16 : REFERENCES 1. W. B. Cotrell, "The ECCS Rule-Making Hearing," Nuclear Safety, Vol. 15, no.1, 1974. 2. John G. Collier and Geoffrey F. Hewitt, “Introduction to Nuclear power,” Hemisphere publishing Corporation, Springer-Verlag, 1987. 3. James H. Rust, “Nuclear power Plant Engineering,” Haralson Publishing Company, 1979. 4. Arthur R. Foster, and Robert L. Wright, Jr, “Basic Nuclear Engineering,” Allyn and Bacon, Inc., 1978. ...
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This note was uploaded on 06/16/2010 for the course NPRE 402 taught by Professor Ragheb during the Spring '08 term at University of Illinois at Urbana–Champaign.

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