Simplified BWR

A design example is a system of 1,700 MWe with an operating pressure of 25 MPa and a reactor outlet temperature of 510°C, which is expected to range up to 550°C (O Fig. 58.5). This feature enables ca. 44% of thermal efficiency, which is about one third higher than current LWRs. Passive safety features are incorporated similar to those of simplified BWRs. 

IX-4] HEKI, H., et al.. Long operating cycle simplified BWR, ICONE-9 (Proc. 9 Int. 

IX-9] YOSHIDA, N., et al., Fuel and core design for long operation cycle simplified BWR (LSBWR) Advanced Nuclear Power Plants (Proc. Int. Congress, Hollywood, FL, USA, June 9-13, 2002), ICAPP 02, No. 1032. 

Figure 6,1-2 is a simplified illustration of a BWR. The pressure of the moderator-cooling water at about half the pressure in a PWR forms steam as it flows upward through the core. Steam passes through a moisture separator (shown as vertical lines just above the core) exits the containment through Main Steam Isolation Valves (MSIV) drives the turbine and generates electricity. After the steam is cooled by the turbine, it is condensed, and pumped back to the reactor by the feedwater pump. 

Having completed the risk analyses, computed the uncertainties, and identified critical systems by importance measures (which also identifies valuable systems improvements having low costs), the PSA results must be presented. An executive summary compares the risk of operations that were analyzed with the risks of similar operations. It identifies and explains the main contributors to the risk to people untrained in PSA and statistical methods. Figure 6.3-5 shows two pie-charts that show the risk contributions of various initiators for PWRs and BWRs. A chart similar to one of these would be an effective way of showing the risk contributions in simplified form. 

The capacities of the emergency core cooling systems suffice to provide water under all postulated pipe break conditions. This statement is also valid assuming that only two of the four redundant subsystems are operable. The postulated loss-of-coolant conditions include a hypothetical 80 cm leak at the bottom of the reactor vessel In this context, it can be noted that the capacity of the low pressure coolant injection punq)S has been reduced for BWR 90, following comprehensive core cooling analyses. As a secondary effect, it has been possible to simplify the auxiliary power supply systems. 

PIUS is a new reactor concept based on well established LWR technology and infrastructure. It is a passive and simplified, reconfigured PWR incorporating also some BWR features, with a nominal power output of600 MWe. The modified arrangement reflects the goals of achieving increased simplicity and safety, in particular with respect to protection of the reactor core in possible accident scenarios. 

A 52" turbine is adopted as model TCDF-52, which had already been developed as TC6F-52 for the 1350 MWe BWR, in order to simplify the turbine island... 

A-8a. K. Fischer and W. Hafher, Retention of Aerosols in Water Pools, BF-V38.070-01, Battelle Engineering Corp., Frankfurt am Main, Germany, March 1994, and A-8b. D.A. Powers, A Simplified Model of Decontamination by BWR Steam Suppression Pools,... 

A small Boiling Water Reactor (BWR) design concept has been developed at GE which maximizes the use of BWR design, technology and operating experience. Modest innovations are included to simplify the performance of safety functions. These, as well as other system simplifications, and a reduced power rating less than 600 MW(e) can reduce total costs and speed construction. 

This small BWR concept (Figure 1) uses an isolation condenser to improve transient response. Gravity-driven control rods and gravity-driven borated water injection are used to simplify and provide diversity to the shutdown function. Core cooling and decay heat removal are provided by depressurizing the reactor to an elevated suppression pool. The drywell and pool gas spaces are inert. 

Steam is produced in the reactor vessel in a manner similar to that of current BWRs. The forced recirculation system of large BWRs is replaced with natural circulation. The steam-water mixture exiting the core is directed to separators and dryers which are positioned above and around the core periphery to allow entry of control rod drives to the top of the core. Control rod drives are mounted on the top head to reduce vessel and building size, to simplify the shutdown system, and to minimize penetrations below the core. 

Figure VIII-1 shows a simplified schematic diagram of the nuclear steam supply system with the Package-Reactor. The concept resembles a calandria-type pressurized heavy water reactor (e.g., the FUGEN advanced thermal reactor (ATR) or CANDU reactors) since all these employ pressure tubes. But the Package-Reactor is somewhat different from the ATR or the CANDU. The Package-Reactor employs natural circulation with two-phase flow for core cooling and has no recirculation pumps. The height of the pressure tubes of the cassettes is required to be as low as possible to attain a compact unit. Two-phase flow with high void fractions similar to BWRs is adopted to attain natural circulation with a cassette height of 6 m and a fuel rod length of 3.65 m.

The case study model used in this work is based on a PRA model of a nuclear power plant. The model depicts a Active and simplified boiling water reactor (BWR) type nuclear power plant (NPP). In this section, the model is only briefly introduced on a general level. Detailed information on the PRA model can be... 

The reactor concept presented herein (the CCR) takes a follow-up on the LSBWR. It has a small power output, the capability of long operating cycles and a simplified and compact BWR type configuration with comprehensive safety features. To be economically competitive, the CCR design includes simplification of systems and compact structure. 

IX-16] H. HEKI et al.. Development of simplified compact containment BWR plant, ICONE-12 ((Proc. 12 Int. Conf on Nuclear Engineering, Arlington, VA USA, April 25-29, 2004), ICONEl2-49113. 

In the same year, one of the annual review articles in Industrial and Engineering Chemistry 2) mentioned that the BWR equation of state seemed to provide the most accurate method thus far developed for estimating K-factors for hydrocarbon systems. Use of the equation was deemed tedious, and a procedure for using the equation in a simplified form suitable for rapid equilibrium calculations was to be presented. Charts based on the procedure were available from the M, W. Kellogg Co., New York. 

Figure A1.18 Simplified iayout of typicai BWR NPP (courtesy of US NRC) geneiai basic features (1) thermal neutron spectrum (2) UO2 fuel (3) fuel enrichment about 3% (4) direct cycie with steam separator (steam generator and pressurizer are eiiminated), ie, singie-flow circuit (singie loop) (5) RPV with verticai fuel rods (elements) assembled in bundle strings cooled with upward flow of fight water (water and water—steam mixture) (6) reactor coolant, moderator, and power cycle working fluid are the same fluid (7) reactor cooiant outlet parameters pressure about 7 MPa and samration temperature at this pressure is about 286°C and (8) power cycle subcritical-pressure regenerative Rankine steam turbine cycie with steam reheat.

In LWRs, buckling collapse and creep rupture are not included in the design failure modes, because experimental verifications have shown that these failure modes are not limiting as long as the plastic deformation of the fuel rod is less than 1.0%. The core pressure and temperature of the Super LWR are much higher than those in LWRs, so these failure modes need to be included in the design failure modes. The evaluations of stresses on the cladding are based on ASME Boiler and Pressure Vessel Code Section III as adopted in BWRs for simplified evaluations. 

The main characteristics of Super LWR plant components were briefly introduced with comparison with conventional PWRs, BWRs, and FPPs. The plant system is compact and greatly simplified from those of LWRs which will contribute to the reduction of plant capital cost. The relatively higher thermal efficiency also ensures high economic performance. 

These models consider either the thermodynamic or mechanical non-equilibrium between the phases. The number of conservation equations in this case are either four or five. One of the most popular models which considers the mechanical non-equilibrium is the drift flux model. If thermal non-equilibrium between the phases is considered, constitutive laws for interfacial area and evaporation/condensation at the interface must be included. In this case, the number of conservation equations is five, and if thermodynamic equilibrium is assumed the number of equations can be four. Well-assessed models for drift velocity and distribution parameter depending on the flow regimes are required for this model in addition to the heat transfer and pressure drop relationships. The main advantage of the drift flux model is that it simplifies the numerical computation of the momentum equation in comparison to the multi-fluid models. Computer codes based on the four or five equation models are still used for safety and accident analyses in many countries. These models are also found to be useful in the analysis of the stability behaviour of BWRs belonging to both forced and natural circulation type. 

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