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In this study, the model of Lungmen ABWR NPP (nuclear power plant) using RELAP5 with the interface SNAP was established. This RELAP5/SNAP model performed the mitigation strategy simulation and study. The main systems of the mitigation strategy are RCIC (Reactor Core Isolation Cooling System), ADS (Automatic Depressurization System), and AC-Independent Water Addition System (ACIWA). This research focuses to assess the usefulness of the mitigation strategy under Fukushima-like conditions. There are three steps in this study. The first step is the establishment of Lungmen NPP RELAP5/SNAP model using the systems data, startup tests reports, and FSAR. The second step is the mitigation strategy simulation and analysis under Fukushima-like conditions using the RELAP5/SNAP model. In this step, the case without mitigation strategy was also performed to evaluate the mitigation strategy effectiveness. Finally, the FRAPTRAN analysis with the interface SNAP was performed in this study. This can evaluate the mechanical property and integrity of fuel rods. According to the RELAP5 results under Fukushima-like conditions, the mitigation strategy can maintain the PCT (peak cladding temperature) below the criteria (1088.7 K). And the FRAPTRAN results also present that the mitigation strategy can keep the fuel rods integrity. It indicates that Lungmen NPP would be in a safe situation when Lungmen NPP performs the mitigation strategy.

After Fukushima Daiichi disaster, the safety analysis of severe accidents became one of the safety concerns in Taiwan. At the same time, Taiwan Power Company built up an event basis strategy called Ultimate Response Guideline (URG) to cope with the accident like Fukushima. The analysis of the effectiveness of URG also became the most popular issue in Taiwan. However, the beyond design basis accident was very complicate and combined with many different aspects. Although researchers in Taiwan did lots of analysis of URG using different codes with different aspects, the differences of codes in the severe accident analysis were not mention in any of those studies. Thus, a Fukushima-like accident of Chinshan Nuclear Power Plant (NPP) was done by four codes in this study: MELCOR2.1, TRACE, PCTRAN and MAAP5.03. All four codes shared the same boundary conditions in the accident simulations. The crosswalk of these codes was discussed in detail and showed the similarity and differences among these codes. Those similarity and differences can help the improvement of URG analysis and make that this kind of analysis has more reliability. Also, the application field of each code can be clearly distinguished and help the improvement of code use in future NPP safety analysis.

After the Fukushima Daichii accidents, Taiwan Power Company developed a strategy to cope with such extended SBO (Station Blackout) cases at nuclear power plants, which called URG. The MAAP and PCTRAN were used to perform the study for Kuosheng BWR/6 nuclear power plant (NPP) ultimate response guideline (URG). The main actions of URG are the depressurization and low pressure water injection of the reactor and the venting of the containment. This study focuses to confirm the URG efficiency. The analysis results depict that following the URG, the fuel can be covered by the coolant, no exposure. It can also prevent the radiation release and the large evacuation. It indicated that Kuosheng NPP was at the safe situation. It shows that the two-step depressurizations can extend the time of the preparation of alternate water source. The minimum injection rate to prevent the fuel to expose is 192 gpm in MAAP.

Maanshan nuclear power plant (NPP) is a two-unit Westinghouse three-loop PWR NPP. This research focuses on the analysis and simulation of Maanshan NPP station blackout (SBO) accident that happened on 18th March, 2001, and thermal–hydraulic phenomena of the plant during SBO with and without mitigation strategies. There are two main steps in this study. The first step was the establishment of Maanshan NPP SNAP/TRACE models. These models were tested and the results of TRACE were compared with the startup test and FSAR data. Analysis results indicate that Maanshan NPP SNAP/TRACE models predict not only the behaviors of important plant parameters consistently with the startup test and FSAR data, but also their associated numerical values with respectable accuracy. The next step was the SBO accident simulation and analysis of Maanshan NPP SNAP/TRACE model. The results of TRACE show good agreement with the plant data. Several mitigation strategies were simulated and studied by using this model in this research.

Chinshan Nuclear Power Plant (NPP) was built in 1970 and is a BWR/4 NPP. The original rated power of Chinshan NPP for each unit is 1775 MW. Chinshan NPP finished the project of SPU (stretch power uprate) and the operating power is 1840 MW now. In order to ensure the NPPs safety, the simulation and analysis of the transients for the NPPs are very important work. After the Fukushima NPP disaster occurred, there is more concern for Taiwan NPPs safety. For the simulation of ultimate response guideline (URG), the TRACE/SNAP model of Chinshan NPP was established in this research. The reactor depressurization, low pressure water injection, and containment venting are the main actions of URG. This research focuses to assess the URG utility of Chinshan NPP under Fukushima-like conditions. This study consists of four steps. The first step is the establishment of Chinshan NPP TRACE/SNAP model. In order to evaluate the TRACE/SNAP model’s system response, startup tests and FSAR data were used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like conditions by using Chinshan NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to confirm the URG effectiveness of Chinshan NPP. Additionally, the MELCOR/SNAP model of Chinshan BWR/4 NPP was also established for the above cases. The prediction comparison of TRACE and MELCOR was done. Third, the sensitivity study of URG and the required raw water injection were performed. According to TRACE results, the URG can keep the peak cladding temperature (PCT) below the criteria 1088.7 K under Fukushima-like conditions. It indicates that Chinshan NPP can be controlled in a safe situation. If Chinshan NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower than TAF (top of active fuel) which means a safety issue about the fuel rods may be generated. Finally, by using FRAPTRAN code and the results of TRACE, the analysis was performed in order to evaluate the mechanical property and integrity of fuel rods.

Lungmen nuclear power plant is the first advanced boiling water reactor (ABWR) plant in Taiwan. According to the most up-to-date schedule, Lungmen Plant will begin its power tests in later 2011. The tests contain (1) loss of feedwater heater test, (2) one feedwater pump trip test, (3) one reactor internal pump trip test, (4) three reactor internal pumps trip test, (5) loss of offsite power and turbine/generator trip test, (6) turbine trip test, (7) load rejection test, (8) fast load winddown test, and (9) reactor full isolation test. The acceptable values of the plant major parameters in all test criteria of each plant power test have been provided by the plant vendor's analysis results. Thus, the verification of the correctness of the vendor results is important. A dual RELAP5 advanced Lungmen Plant specific simulation platform (ALPS) has been used to simulate the plant power tests. ALPS is developed based on two independent RELAP5 modules: one for the reactor system modeling and the other one for the balance of plant system modeling. All plant systems simulation models are synchronized on a PC-based commercial simulation platform, namely 3KeyMaster. The plant major control systems and the emergency core cooling system are simulated in great detail with the built-in modeling tools of 3KeyMaster. ALPS calculation results have been compared with the plant vendor's preliminary power tests analysis results. The comparison results have been delivered to the plant vendee for references. In this article some typical power tests simulation and comparison results are presented for (1) one feedwater pump trip test, (2) one reactor internal pump trip test, (3) load rejection test, and (4) reactor full isolation test.

Protecting the public from radiation exposure is important if a nuclear power plant (NPP) accident occurs. Deciding appropriate protective actions in a timely and effective manner can be fulfilled by using an effective accident evaluation tool. In our earlier work, we have integrated PCTRAN (Personal Computer Transient Analyzer) with the off-site dose calculation model. In this study, we introduce PCTRAN as an evaluation tool for nuclear power plant emergency responses. If abnormal conditions in the plant are monitored or observed, the plant staffs can distinguish accident/incident initiation events. Thus, the responsible personnel can immediately operate PCTRAN and set up those accident/incident initiation events to simulate the nuclear power plant transient or accident in conjunction with off-site dose distributions. The evaluation results consequently help the responsible organizations decide the rescue and protective actions. In this study, we explain and demonstrate the capabilities of PCTRAN for nuclear emergency responses, through applying it to simulate the postulated nuclear power plant accident scenarios.

The modular accident analysis program (MAAP) is a fast-running severe accident analysis tool with which the timing of key events and source terms in a severe accident are assessed. The idea of combining MAAP and an optimization algorithm to identify the realistic accident parameters in terms of minimizing the discrepancies between the plant data and the simulation results is straightforward. In 2008 Chien and Wang first compiled the combination of the MAAP4 source codes and a Simplex code as a computer-aided tool for the loss-of-coolant accident (LOCA) of the Kuosheng nuclear power plant (NPP). The break area and break elevation were successfully identified. However, in that approach to putting the idea into practice was that hard data dependence exists between MAAP and the optimization algorithm. Tedious tracing and modification work is required to ensure all plant variables in MAAP source codes with the exception of the adjusted accident parameters are identical at the beginning of every simulation. The plant- and accident-specific development features also easily limit the applications of this idea to the nuclear industry, like being boxed in. In this study a so-called "out-of-box" approach is proposed that can omit the limits of the idea applications on severe accident management. A parameter identification tool developed in this approach for the same postulated LOCA of the Kuosheng NPP is carried out for verification and validation. It demonstrates the advantages of successful parameter identification, less programming efforts, and no plant- and accident-specific features.

The minimum number of safety/relief valves (SRVs) required for emergency depressurization (MNSRED) is an important reactor pressure vessel (RPV) parameter described in the appendix C of BWR Owners’ Group, Emergency Procedure Guidelines and Severe Accident Guidelines (BWROG, EPG/SAGs). It is directly related to the emergency depressurization if there are less SRVs available. MNSRED is always coupled with decay heat removal pressure (DHRP) in emergency depressurization. DHRP is used as the depressurization low-pressure end to ensure the decay heat can be removed. The purpose of this paper is to discuss their mathematical models in BWROG EPG/SAGs. After study, the vague concepts existed in these parameters are pointed out and clarified. The variables inappropriately used through their evaluations are revised. With improved modifications, the MNSRED turns out smaller than its original value and DHRP is larger than its previous one. These results affect the heat removal capability and the operation margin regarding emergency depressurization. At the end, the revised DHRP value is verified by MAAP5 code with the same plant status for the Kuosheng power station of Taiwan Power Company.

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.

Generic severe accident guidelines (SAGs) of a boiling water reactor are developed based on the Mark-I containment system. There are many intrinsic features of Mark-III containment including the following: the drywell does not form a part of the primary containment boundary, it has large containment free volume, the drywell is enclosed by the primary containment, and the drywell and the primary containment are located parallel on the same elevation. Because of the features of the Mark-III system and its benefit to the containment flooding strategy, it is found that the associated containment flooding strategy can be simplified. The seven legs in the containment flooding strategy are reduced to four legs. The five rows of check logic are reduced to three. Simplifying the SAG charts whenever possible increases the probability of a successful outcome. The tedious and time-consuming work of identifying the reactor pressure vessel status can be removed. It saves time for decision making in emergency conditions.

The MAAP4 code is a fast-running severe accident analysis and management code widely used in nuclear industrial applications. It has been used as a severe accident analysis tool with which the timing of key events and source terms in the accident are assessed in Taiwan for many years. In the case of a loss-of-coolant accident (LOCA), it needs to know its break area and break elevation prior to reconstructing the accident for further analysis. For the system with more than one unknown parameter, it is not easy to identify these parameters since their tuning work becomes sophisticated. The purpose of this paper is to find a better solution for this problem by coupling an optimization module with the MAAP4 simulation code. In this paper, a recirculation line break LOCA of the Kuosheng nuclear power plant is cited as a reference case. Different recirculation line break areas are tested and verified. This optimization module coupled with the MAAP4 code can be used to identify the break elevation and area if the shroud water level drops below the break point.

A study of the in-phase (global, core-wide) and out-of-phase (regional) modes of stability with LAPUR5 code for Kuosheng NPP is presented. Comparisons made with vendor’s STAIF results indicated close agreements, within the acceptable 20% deviation in decay ratios, for Kuosheng NPP Unit 1 Cycle 18 reloads design. Parametric studies for stability have shown that the inlet loss coefficient, exposure, control rod position, and axial power shape are the most important parameters that might affect the accuracy of decay ratios for in-phase and out-of-phase modes of stability analysis. A decrease in the inlet loss coefficient or increase in the control rods insertion causes an increase in the global and regional decay ratios. An increase in the exposure causes an increase in the global decay ratio, the maximum value of the regional decay ratio occurs in MOC condition, and the regional decay ratio is higher than global decay ratio in the BOC condition.

Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The execution of containment flooding requires information about the water level in the primary containment. However, there is no instrument to measure the drywell water level for most Mark-III systems. Furthermore, because of the design feature of the Mark-III containment, the water level in the containment does not necessarily guarantee that there is an equivalent water level in the drywell. Therefore, the development of a drywell water level computational aid becomes very useful. The purpose of this work is to develop and validate the drywell water level computational aid and to investigate the implementation of the proposed computational aid on the containment flooding strategy of a Mark-III system. The Kuosheng nuclear power plant (NPP) is a typical BWR-6 NPP with Mark-Ill containment, and the Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners' Group Emergency Procedure Guidelines and Severe Accident Guidelines, Revision 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code is chosen as the tool for analysis.

Containment flooding is an important strategy for severe accident management of a conventional boiling water reactor (BWR) system. The purpose of this work is to investigate the containment flooding strategy of the Mark-III system after a reactor pressure vessel (RPV) breach. The Kuosheng Power Plant is a typical BWR-6 nuclear power plant (NPP) with Mark-III containment. The Severe Accident Management Guideline (SAMG) of the Kuosheng NPP has been developed based on the BWR Owners Group (BWROG) Emergency Procedure and Severe Accident Guidelines, Rev. 2. Therefore, the Kuosheng NPP is selected as the plant for study, and the MAAP4 code is chosen as the tool for analysis. A postulated specific station blackout sequence for the Kuosheng NPP is cited as a reference case for this analysis. Because of the design features of Mark-III containment, the debris in the reactor cavity may not be submerged after an RPV breach when one follows the containment flooding strategy as suggested in the BWROG generic guideline, and the containment integrity could be challenged eventually. A more specific containment flooding strategy with drywell venting after an RPV breach is investigated, and a more stable plant condition is achieved with this strategy. Accordingly, the containment flooding strategymore » after an RPV breach will be modified for the Kuosheng SAMG, and these results are applicable to typical Mark-III plants with drywell vent path.« less

The work described in this paper is benchmark calculations of pressurization transient turbine trip tests performed at the Peach Bottom boiling water reactor (BWR). It is part of an overall effort in providing qualification basis for the INER/TPC BWR transient analysis method developed for the Kuosheng and Chinshan plants. The method primarily utilizes an advanced system thermal hydraulics code, RETRAN02/MOD5, for transient safety analyses. Since pressurization transients would result in a strong coupling effect between core neutronic and system thermal hydraulics responses, the INER/TPC method employs the one-dimensional kinetic model in RETRAN with a cross-section data library generated by the Studsvik-CMS code package for the transient calculations. The Peach Bottom Turbine Trip (PBTT) tests, including TT1, TT2, and TT3, have been successfully performed in the plant and assigned as standards commonly for licensing method qualifications for years. It is an essential requirement for licensing purposes to verify integral capabilities and accuracies of the codes and models of the INER/TPC method in simulating such pressurization transients. Specific Peach Bottom plant models, including both neutronics and thermal hydraulics, are developed using modeling approaches and experiences generally adopted in the INER/TPC method. Important model assumptions in RETRAN for the PBTT test simulations aremore » described in this paper. Simulation calculations are performed with best-estimated initial and boundary conditions obtained from plant test measurements. The calculation results presented in this paper demonstrate that the INER/TPC method is capable of calculating accurately the core and system transient behaviors of the tests. Excellent agreement, both in trends and magnitudes between the RETRAN calculation results and the PBTT measurements, shows reliable qualifications of the codes/users/models involved in the method. The RETRAN calculated peak neutron fluxes of the PBTT tests are higher than the measurements, which ensures the conservatism of using the INER/TPC method for Kuosheng and Chinshan licensing safety analyses.« less

Reactor coolant system (RCS) injection is an important strategy for severe accident management of a pressurized water reactor (PWR) system. Maanshan is a typical Westinghouse PWR nuclear power plan...

The current emergency operating procedures (EOPs) for a three-loop pressurized water reactor on a total loss-of-feedwater (LOFW) incident are experimentally evaluated at the Institute of Nuclear Energy Research Integral System Test (IIST) facility. Four experiments were conducted with a different number of pressurizer power-operated relief valves (PORVs) and centrifugal charging pumps (CCPs) engaged in performing the primary-side bleed-and-feed operation. In addition, the effect of opening a pressure vessel venting-path was also evaluated. The focus was on the effectiveness and adequacy of current EOPs and the observation of associated thermal-hydraulic phenomena. The test results indicated that the current EOPs can function well in mitigating the consequences of a LOFW event provided that at least two fully opened PORVs and one actuated CCP are used. The opening of a pressure vessel venting path during the bleed-and-feed operation is helpful in the primary system depressurization but has little effect to increase the pressure vessel coolant inventory. However, by opening one PORV, precaution for the reactor coolant system (RCS) integrity should be taken for system overpressurization caused by insufficient heat removal capability. Furthermore, an effective control of restored feedwater to a low injection rate is favorable to minimize the thermal stress in the RCSmore » pressure boundary.« less

The current emergency operating procedures (EOPs) for a three-loop pressurized water reactor on steam generator tube rupture (SGTR) incident are experimentally evaluated at the Institute of Nuclear Energy Research Integral System Test Facility. The focus is on the adequacy of EOPs to limit primary-to-secondary leakage with particular emphasis on the number of ruptured U-tubes on the severity of an incident and the response time available for operator actions. By comparing test data with plant records of a Mihama-2 SGTR event, the key thermal-hydraulic phenomena during SGTR transients can be successfully simulated. The test results indicated that the current EOPs can function well in terminating the break flow and maintaining adequate core cooling. However, the effectiveness in minimizing the radioactive release demands an early and substantial operator involvement. To mitigate the consequences of such an event, the timing to isolate the faulted steam generator (SG) and to terminate safety injection flow will be critical. Furthermore, to avoid overfilling, the water level in the faulted SG needs to be drained prior to the implementation of the cooldown process.

Three analyses are presented for the Kuosheng plant, the second nuclear power station of the Taiwan Power Company, The first is a startup test benchmark analysis of full closure of the main steam isolation valves (MSIVs ). This analysis is one ofa series of startup test analyses to develop the Kuosheng RETRAN model. The second is a parallel reload safety analysis of a system pressure-limiting transient for Kuosheng Unit 2, Cycle 12, which is a part of work to develop an in-house reload safety analysis methodology, The third is an operating support analysis of a reactor trip event due to a single MSIV closure in October 1997; this analysis demonstrates the application of the Kuosheng RETRAN model. The work may add to the RETRAN experience base and to the plant modeling body of knowledge. A similar approach can be adopted to support the plant, with concerns.

The RETRAN analyses of three areas of thermal-hydraulic concerns are presented. The first evaluated whether a 360-deg through-wall crack could be detected based on system response with leakage through the crack. The second analysis was to identify the hydraulic load on the shroud and possible consequences early in the main steam-line break. The last was the evaluation of the thermal limit (ACPR) oflimiting transients with leakage. Some of the analyses were used for justification of continued operation of the current cycle, and some will be useful in the evaluation of the repair design.

Two analyses for Chinshan nuclear power station (CSNPS) show the RETRAN applications in reactor operations. One was used to find the root cause of a reactor trip due to high neutron flux (APRM Hi-Hi). Two of the four turbine control valves failed and closed rapidly. Whether the pressure increase caused the APRM Hi-Hi reactor scram was questioned. An analysis was performed to determine the way to bring unit I back to power. The second analysis was conducted to provide an evaluation of the consequences before a plant test planned for late 1994 to consider a design change upgrading recirculation flow control. However, there were concerns about reactor scram during execution of the test and the potential of entering the exclusive region of instability. Sensitivity studies on several different combinations of initial conditions and feedwater flow coastdown curves indicated potential problems with reactor trip and stability.