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The limiting blowdown event for the design of an advanced boiling water reactor (ABWR) containment shifts from a conventional recirculation line break to a feedwater line break (FWLB) by implementing reactor internal pumps. As a result, coupled blowdown from both the reactor pressure vessel (RPV) and the balance of plant (BOP) is involved in the limiting FWLB. Coupled blowdown from both RPV and BOP for the FWLB of the Lungmen ABWR has been successfully analyzed using the advanced RELAP5-3D/K code. To simulate adequately both the RPV and BOP blowdown, the essential simulation scope of an ABWR includes the reactor system, the main steam and turbine systems, the condensate and feedwater systems, the protection system, and the emergency core cooling system. As compared to what was presented in the preliminary safety analysis report of the Lungmen ABWR, unexpected prolonged decays of BOP blowdown flow and enthalpy were calculated. The revised blowdown flow and enthalpy calculated by RELAP5-3D/K from both RPV and BOP breaks provide a new and solid basis for the final safety analysis of ABWR containment for the Lungmen plant, which is scheduled for commercial operation in 2011. The successful modeling of the entire RPV and BOP with RELAP5-3D/K and associated application to the FWLB licensing blowdown analysis indicate that the advanced RELAP5 code can extend its traditional reactor safety analysis to the simulation and analysis of the entire power generation and conversion systems.

In the innovative design of the advanced boiling water reactor (ABWR), conventional recirculation loops are removed and replaced by multiple reactor internal pumps. Therefore, there is no major penetration of the reactor pressure vessel (RPV) below the elevation of the top of active fuel. As a result, an ABWR loss-of-coolant accident (LOCA) can have a decreased impact on reactor safety. Moreover, in the new RPV design the injection points of all the conventional low-pressure emergency core cooling (ECC) systems (ECCSs) are shifted out of the core shroud to the downcomer andfeedwater line as a new low-pressure ECCS, namely, a low-pressure flooder (LPFL). Consequently, the net hydraulic head built inside the downcomer will be the only driving force to bring the low-pressure ECC water into the core shroud during a large-break LOCA. In the analysis of a feedwater line break with RELAP5-3D/K, it was occasionally found that the hydraulic head built in the downcomer might not be great enough to bring the ECC water into the core shroud, and when the mixture water column ascends above the elevation of the feedwater rings, all the water injected by the LPFL will be directly driven to the break on the feedwater line. Fortunately, the capacity of the remaining high-pressure ECC flow directly injected above the core is great enough, and this ECC low-pressure injection bypass phenomenon can be terminated once the high-pressure ECC injection is manually turned off. This phenomenon of low-pressure ECC injection bypass is unexpected in the ABWR design, and it is worth further investigation.

The blowdown of feedwater (FW) line breaks (FWLBs) has been successfully analyzed by using the Appendix K version of RELAP5-3D. To adequately simulate a feedwater blowdown event, one must consider the main steam system, the turbine system, the moisture separator reheaters (MSRs), the main condenser, and the condensate and FW system as all are involved in the modeling scope. The essential components of the simulation scope include the steam header, the high-and low-pressure turbines, the MSR, the FW pump (FWP) turbines, the main condenser, the condensate and booster pumps, the FW heaters of six stages, the steam extraction of seven stages, and the turbine-driven FWPs. All of the components are connected by associated piping as designed. Regarding the FW blowdown analysis, blowdown mass and energy are the two most important parameters to be calculated. Several essential phenomena are involved in this FW blowdown event, which include critical flow at the break and the internal venturi, flashing of FW near the break, runout and coastdown of the FWPs, steam extraction to FW heaters and FWP turbines, flashing of saturated water initially stored inside the FW heater shell sides and MSR drain tanks, energy release from saturated water and system metal, and cold water transportation from the main condenser to the break. All the essential processes involved during FWLB can be well simulated by the advanced Appendix K version of RELAP5-3D. The blowdown analysis calculated by RELAP5-3D/Kfor the FWLB was contracted to provide a solid basis for the final safety analysis report containment design analysis for the Lungmen advanced boiling water reactor (ABWR) plant. The successful application of RELAP5 for the entire balance-of-plant simulation and associated FW blowdown analysis indicates that the advanced RELAP5 can extend its traditional reactor safety analysis to the entire power conversion system simulation and analysis.

In light water reactors, particularly the pressurized water reactor (PWR), the severity of a loss-of-coolant accident (LOCA) would limit how high the reactor power can operate. Although the best-estimate LOCA licensing methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during a LOCA, it generally takes much more resources to develop. Instead, implementation of evaluation models required by Appendix K of 10CFR50 on an advanced thermal-hydraulic platform such as RELAP5, TRAC, etc., also can gain significant margin for the PCT calculation. Through compliance evaluation against Appendix K of 10CFR50, all of the required evaluation models have been implemented in RELAP5-3D. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effects experiments and eight sets of LOCA integral experiments were adopted. Through the assessments against separate-effects experiments, the success of the code modification in accordance with Appendix K of 10CFR50 was demonstrated. Besides, one set of a typical integral large-break LOCA from Loss-of-Fluid Test Facility experiments (L2-5) has also been applied to preliminarily evaluate the integral performance of the Appendix K version of RELAP5-3D. The PCT predicted by the evaluation models is greater than the one from best-estimate calculation inmore » the whole LOCA history with the conservatism of 150 K, and the measured PCTs of L2-5 are also well bounded by the evaluation model calculation. Another seven sets of integral-effect experiments will be further applied in the next step to ensure the reasonable integral conservatism of the newly developed LOCA licensing analysis code (RELAP5-3DK/INER), which can cover all the phases of both large- and small LOCA in one code.« less

With the consideration of mass unbalance, coolant shrinking, and compressibility, a model for reactor coolant leakage evaluation has been developed to quantify on-line the system leakage rate with conventional system measurements, regardless of where the leak occurs. This model has been derived from the system of total continuity, and it divides the reactor coolant system (RCS) into two regions, namely, the saturated and subcooled regions. The pressurizer is considered as a saturated region, and the remaining part of the RCS is regarded as a subcooled region. Taking the on-line measurements of the RCS including the RCS pressure, temperature, pressurizer water level, and charging and letdown flow rates, this model can directly evaluate on-line the RCS leakage rate. It is noted that this model is applicable only if the RCS remains subcooled. To verify the applicability of this model, data generated by RELAP5/MOD3 simulation and experimental measurements from the Institute of Nuclear Energy Research, Taiwan, Integral System Test Facility were adopted to assess this model. With further on-line verification against the Maanshan training simulator, this model was finally delivered to the Maanshan nuclear power plant (a three-looped Westinghouse pressurized water reactor) to assist the operator training and on-line evaluation of themore » RCS leakage rate. The smallest amount of leak flow that can be detected by the ROCK model is 3 gal/min.« less

The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical model in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has beenmore » successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future.« less

In light water reactors, particularly the pressurized water reactors, the severity of loss-of-coolant accidents (LOCAs) will limit how high the reactor power can extend. Although the best-estimate LOCA methodology can provide the greatest margin on the peak cladding temperature (PCT) evaluation during LOCA, it will take many more resources to develop and to get final approval from the licensing authority. Instead, implementation of evaluation models required by Appendix K of the Code of Federal Regulations, Title 10, Part 50 (10 CFR 50), upon an advanced thermal-hydraulic platform can also gain significant margin on the PCT calculation. A program to modify RELAP5-3D in accordance with Appendix K of 10 CFR 50 was launched by the Institute of Nuclear Energy Research, Taiwan, and it consists of six sequential phases of work. The compliance of the current RELAP5-3D with Appendix K of 10 CFR 50 has been evaluated, and it was found that there are 11 areas where the code modifications are required to satisfy the requirements set forth in Appendix K of 10 CFR 50. To verify and assess the development of the Appendix K version of RELAP5-3D, nine kinds of separate-effect experiments and six sets of integral-effect experiments will be adopted.more » Through the assessments program, all the model changes will be verified.« less

Although only a few percent of residual power remains during plant outages, the associated risk of core uncovery and corresponding fuel overheating has been identified to be relatively high, partic...