Future directions in boiling water reactor design

The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990s. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste.The ABWR incorporates the best proved features from BWR designs in Europe, Japan, and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced turbine/generator with 52 in. last stage buckets; and advanced radwaste technology.The ABWR is being developed as the next generation Japan standard BWR under the guidance and leadership of the Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. During 1987, the Tokyo Electric Power Company, Inc. announced its decision to proceed with two ABWR units at its Kashiwazaki-Kariwa Nuclear Power Station, with commercial operation of the first unit in 1996 and the second unit in 1998. The units will be supplied by a joint venture of General Electric, Hitachi and Toshiba, with General Electric selected to supply the nuclear steam supply systems, fuel and turbine/generators. In the United States it is being adapted to the needs of U.S. utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the U.S. Nuclear Regulatory Commission for certification as a preapproved U.S. Standard BWR under the U.S. Department of Energy's ALWR Design Verification Program. These cooperative Japanese and U.S. Programs are expected to establish the ABWR as a world class BWR for the 1990s.International cooperative efforts are also underway aimed at development of a simplified BWR employing natural circulation and passive safety systems. This BWR concept, while only in the conceptual design stage, shows significant technical and economic promise.     

Design and test of hydraulic vacuum breaker check valve for simplified boiling water reactor

In the current design of the simplified boiling water reactor, the vacuum breaker check valve is an important safety component. The vacuum breaker check valve is the only key safety components which is not passive in nature. Failure of this mechanical valve drastically reduces the passive containment cooling system cooling capability and hence containment pressure may exceed the design pressure. To eliminate this problem novel vacuum breaker check valve was developed to replace the mechanical valve. This new design is based on a passive hydraulic head, which is fail-safe and is truly passive in operation. Moreover this new design needs only one additional tank and one set of piping each to the wetwell and drywell. This system is simple in design and hence is easy to maintain and to qualify for operation. The passive vacuum breaker check valve performance was first evaluated using RELAP5. Then the passive vacuum breaker check valve was constructed and implemented in the PUMA integral test facility. Its performance was studied in a large break loss of coolant accident simulation test performed in PUMA facility.     

The pressurization transient analysis for Lungmen advanced boiling water reactor using RETRAN-02

A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using Lungmen startup test data.     

Spectral shift rod for the boiling water reactor

A Boiling Water Reactor core concept has been proposed using a new fuel component called spectral shift rod (SSR). The SSR is a new type of water rod in which a water level is formed during core operation. The water level can be controlled by the core recirculation flow rate. By using SSRs, the reactor can be operated with all control rods withdrawn through the operation cycle as well as that a much larger natural uranium saving is possible due to spectral shift operation than in current BWRs. The steady state and transient characteristics of the SSRs have been examined by experiments and analyses to certify the feasibility. In a reference design, a four times larger spectral shift width as for the current BWR has been obtained.     

Concept for monitoring gaseous and aerosol-bound radioactive discharges from boiling and pressurized water reactor plants in the event of design basis and severe accidents

A new concept for monitoring radioactive discharge has been developed. The resulting system presented here is intended to meet the requirements set forth in the German Nuclear Safety Standard KTA 1503.2 (draft) for accident surveillance while at the same time being suitable for activity monitoring during containment venting. The system and typical modes of system operation are described for plants equipped with pressurized water reactors (PWRs) and plants equipped with boiling water reactors (BWRs). A combination of different methods of evaluation allows the space needed for instrumentation as well as the effort required for testing to be minimized.     

Fuel lattice design in a boiling water reactor using an ant-colony-based system

This paper presents a new approach to deal with the boiling water reactor radial fuel lattice design. The goal is to optimize the distribution of both, the fissionable material, and the reactivity control poison material inside the fuel lattice at the beginning of its life. An ant-colony-based system was used to search for either: the optimum location of the poisoned pin inside the lattice, or the U²³⁵ enrichment and Gd₂O₃ concentrations. In the optimization process, in order to know the parameters of the candidate solutions, the neutronic simulator CASMO-4 transport code was used. A typical 10 × 10 BWR fuel lattice with an initial average U²³⁵ enrichment of 4.1%, used in the current operation of Laguna Verde Nuclear Power Plant was taken as a reference. With respect to that reference lattice, it was possible to decrease the average U²³⁵ enrichment up to 3.949%, this obtained value represents a decrease of 3.84% with respect to the reference U²³⁵ enrichment; whereas, the k-infinity was inside the ±100 pcm’s range, and there was a difference of 0.94% between the local power peaking factor and the lattice reference value. Particular emphasis was made on defining the objective function which is used for making the assessment of candidate solutions. In a typical desktop personal computer, about four hours of CPU time were necessary for the algorithm to fulfill the goals of the optimization process. The results obtained with the application of the implemented system showed that the proposed approach represents a powerful tool to tackle this step of the fuel design.     

Automatic boiling water reactor loading pattern design using ant colony optimization algorithm

An automatic boiling water reactor (BWR) loading pattern (LP) design methodology was developed using the rank-based ant system (RAS), which is a variant of the ant colony optimization (ACO) algorithm. To reduce design complexity, only the fuel assemblies (FAs) of one eight-core positions were determined using the RAS algorithm, and then the corresponding FAs were loaded into the other parts of the core. Heuristic information was adopted to exclude the selection of the inappropriate FAs which will reduce search space, and thus, the computation time. When the LP was determined, Haling cycle length, beginning of cycle (BOC) shutdown margin (SDM), and Haling end of cycle (EOC) maximum fraction of limit for critical power ratio (MFLCPR) were calculated using SIMULATE-3 code, which were used to evaluate the LP for updating pheromone of RAS. The developed design methodology was demonstrated using FAs of a reference cycle of the BWR6 nuclear power plant. The results show that, the designed LP can be obtained within reasonable computation time, and has a longer cycle length than that of the original design.     

Heat transfer modeling in the vertical tubes of the passive containment cooling system of the simplified boiling water reactor

The long term containment cooling of GE's passive BWR design is based on a new safety system called PCCS (passive containment cooling system). Performance of this system relies on the pressure difference between the drywell and wetwell in case of an accident and on the condensation of steam moving downward inside vertical tubes fully submerged in a water pool initially at room temperature. In this paper a model based on the resolution of momentum equations of both phases, the application of the heat and mass transfer analogy, and the consideration of the presence of a noncondensable gas by diffusion theory in a boundary layer is presented. Assumptions and approximations taken resulted in new methods to estimate film thickness and heat transport from the gas to the interface. Influence of phenomena such as suction, flow development, film waviness, and droplet entrainment has been accounted for. Based on this formulation, a computer programme called HVTNC (heat transfer in vertical tubes with noncondensables) has been built up. HVTNC results have been compared to the experimental data available. Experimental trends have been reproduced. Heat transfer has been found to be severely degraded by the presence of noncondensables whereas high Reynolds numbers of gas flow have been seen to enhance shear stress and therefore, heat transmission. The average error of HVTNC is essentially located at regions where only a residual fraction of heat remains to be transferred, so that minor deviations can be anticipated in the overall heat transfer in the tube. Comparison of HVTNC to other models show a substantial gain of accuracy with respect to earlier models.     

PbBi-cooled direct contact boiling water small reactor

A design concept of PbBi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. Water is injected into hot PbBi above the core, and direct contact boiling takes place in chimneys. Boiling bubbles rise due to buoyancy effects, which works as a lift pump for PbBi circulation. The generated steam passes through separators and dryers for the removal of PbBi droplets, and then flows into turbines for the generation of electricity. The system pressure of 7 MPa is as the same as that of the conventional boiling water reactors (BWRs). The outlet steam is superheated by 10°C to avoid the accumulation of condensate on a PbBi free surface in the reactor vessel. The control rods are inserted from above, which is different from the original concept. This insertion was chosen since the seal of steam at the top of the reactor vessel is technically much easier than the seal of PbBi at the bottom of the reactor vessel. The electric power of 150 MWe may be the maximum which is practically possible as a small reactor with economic competitiveness to conventional LWRs. A two-region core is designed. A decrease in reactivity was estimated to be 1.5%dk/kk′ for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The design limit of cladding temperature is specified to be 650°C for compatibility of cladding material with PbBi. As a result, the PbBi core outlet temperature becomes 460°C. The PbBi temperature rise in the core is 150°C. The conditions of the secondary coolant steam are as the same as those of conventional BWRs with thermal efficiency of 33%. The core is designed to have the breeding ratio of 1.1 and the refueling interval of 15 years as a reactor with a long-life core. Direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed.     

An interface redesign for the feed-water system of the advanced boiling water reactor in a nuclear power plant in Taiwan

A well-designed human-computer interface for the visual display unit in the control room of a complex environment can enhance operator efficiency and, thus, environmental safety. In fact, a cognitive gap often exists between an interface designer and an interface user. Therefore, the issue of the cognitive gap of interface design needs more improvement and investigation. This is an empirical study that presents the application of an ecological interface design (EID) using three cases and demonstrates that an EID framework can support operators in various complex situations. Specifically, it analyzes different levels of automation and emergency condition response at the Lungmen Nuclear Power Plant in Taiwan. A simulated feed-water system was developed involving two interface styles. This study uses the NASA Task Load Index to objectively evaluate the mental workload of the human operators and the Situation Awareness Rating Technique to subjectively assess operator understanding and response, and is a pilot study investigating EID display format use at nuclear power plants in Taiwan. Results suggest the EID-based interface has a remarkable advantage over the original interface in supporting operator performance in the areas of response time and accuracy rate under both normal and emergency situations and provide supporting evidence that an EID-based interface can effectively enhance monitoring tasks in a complex environment.     

Optimal control of a boiling water nuclear reactor

Application of optimal control to a boiling water nuclear reactor is the theme of this paper. The optimal control problem of a linearized model of a reactor is treated as a regulator problem and feedback control laws are derived to drive the system to steady state in the presence of disturbances. The weighting matrices in the performance index of the regulator problem are suitably changed to yield acceptable closed-loop responses for specific disturbances. The disturbances considered are (i) impulse change in temperature of water at inlet to plenum chamber and (ii) step change in throttle valve area. Then the feedback control laws are implemented on the nonlinear model to illustrate their effectiveness both for large and small disturbances.     

Mechanistic modeling for ring-type boiling water reactor fuel spacer design (3) run-off effect and model formulation

The fuel spacer is one of the components of a fuel rod bundle and its role is to maintain an appropriate rod-to-rod clearance. The fuel spacer influences the liquid film flow distribution in the fuel rod bundle, so that the spacer geometry has a strong effect on thermal hydraulic characteristics of BWR such as critical power and pressure drop in the fuel bundle. In this paper, liquid film flow characteristics were experimentally investigated in a circular channel with ring-type spacers using air and water as test fluids in order to be compared with the analytical results introducing the mechanistic spacer model. The spacer model was composed of three effects; the drift flow effect, the narrow channel effect and the run-off effect. The drift flow effect and the narrow channel effect were discussed in the previous works and the run-off effect is done in this paper. This paper shows the formulation around the spacer with use of the three effects. The proposed model explains well the experimental results of liquid film flow rates and thickness carried out in reference to the spacer thickness, the gap between the channel wall and the spacer.     

Age-related degradation of boiling water reactor vessel internals

Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection.     

Design of the advanced accumulator for the pressurized water reactor

With the increased requirement for nuclear power generation as an effective countermeasure against global warming, Mitsubishi has developed the advanced pressurized water reactor (APWR) by adopting a new component of the emergency core cooling system (ECCS), a new instrumentation and control system, and other newfound improvements. The ECCS introduces a new passive component called the advanced accumulator which integrates both functions of the conventional accumulator and the low-pressure pump without any moving parts. The advanced accumulator uses a new fluidics device that automatically regulates flow rates of injected water in case of a loss of coolant accident (LOCA). This fluidics device is referred to as a flow damper. This paper describes the design method of the flow damper and the standpipe.     

Study and application of boiling water reactor jet pump characteristic

RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model.In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible.In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K_d, the suction flow junction form loss coefficients, K_S, the diffuser form loss coefficient, K_e, and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve.     

Dissolution behavior of core structure materials by molten corium in boiling water reactor plants during severe accidents

For the decommissioning of the Fukushima Daiichi Nuclear Power Plant, it is necessary to consider the access route to the fuel debris for its removal, which can be determined by knowing the corruption situation of the core support structure. To predict the damage condition of reactor vessel, dissolution behavior of the core structure material should be understood. In this study, the dissolution behavior of core structure materials (stainless steel) by molten metallic corium (stainless steel + B₄C) originated from control rod and its cladding was investigated. As a result of immersion experiment, it was found that there were two types of dissolution mode in this system: (1) chemical dissolution by eutectic reaction between Fe and B and (2) physical dissolution caused by the grains falling off from solid steel due to infiltration of molten metal. Moreover, on the basis of kinetic analysis, it was considered that the chemical dissolution in this system was slow. Therefore, the dissolution is considered to mainly occur through the mechanism that physical dissolution precedes chemical dissolution.     

Recycling heterogeneous americium targets in a boiling water reactor

One of the limiting contributors to the heat load constraint for a long term spent fuel repository is the decay of americium-241. A possible option to reduce the heat load produced by Am-241 is to eliminate it via transmutation in a light water reactor thermal neutron environment, in particular, by taking advantage of the large thermal fission cross section of Am-242 and Am-242m. In this study we employ lattice loading optimization techniques to define the loadings and arrangements of fuel pins with blended americium and uranium oxide in boiling water reactor bundles, specifically, by defining the incineration of pre-loaded americium as an objective function to maximize americium transmutation. Subsequently, the viability of these optimized lattices is tested by assembling them into bundles with Am-spiked fuel pins and by loading these bundles into realistic three-dimensional BWR core-wide simulations that model multiple reload cycles and observe standard operational constraints. These simulations are possible via our collaboration with the Westinghouse Electric Co. which facilitates the use of industrial-caliber design tools such as the PHOENIX-4/POLCA-7 sequence and the Core Master 2^© GUI work environment for fuel management.     

Analyses of pressure perturbation events in boiling water reactor

In order to design more stable and safer core configurations, experimental and theoretical studies about BWR (Boiling Water Reactor) instability have been performed to characterize the phenomenon and to predict the conditions for its occurrence. The instabilities can be caused by interdependencies between thermal-hydraulic and reactivity feedback parameters such as the void-coefficient, for example, during a pressure perturbation event. In this work, the RELAP5-MOD3.3 thermal-hydraulic system code and the PARCS-2.4 3D neutron kinetic code were coupled to simulate BWR transients. The pressure perturbation is considered in order to study in detail this type of transient. Two different algorithms developed at the University of Pisa were used to calculate the Decay Ratio (DR) and the natural frequency (NF) from the power oscillation signals obtained from the transient calculations. The validation of a code model set up for the Peach Bottom-2 BWR plant is performed against Low-Flow Stability Tests (LFST). The four series of Stability Tests were performed at Peach Bottom Unit 2 in 1977 at the end of cycle 2 in order to measure the reactor core stability margins at the limiting conditions used in design and safety analysis.