[en] This paper reports on a turbine trip test performed at the CHIN-SHAN power station unit 1 that was analyzed using the TPC plant analyzer. The TPC plant analyzer was converted from the Brookhaven National Laboratory plant analyzer and modified for the CHIN-SHAN power station. Comparison of the calculated system response to the plant data indicates that the TPC plant analyzer reproduces the trends of this transient very well. The sequence of events is also close to that of test result

[en] The Taiwan Power Company (TPC) has elected in recent years to implement the power uprate program as a key measure to improve the performance for TPC's nuclear power plants. The Measurement Uncertainty Recapture (MUR) power uprate for the TPC's three operating plants (reported in 16th PBNC) had been successfully implemented by July 2009. For the stretch power uprate (SPU) followed, the magnitude of uprate (~3%) is determined based on the available margins for original plant design, constant pressure approach (BWR) is adopted to simplify the evaluation, and major plant modifications are not considered. As the first application, the SPU safety analysis report (SAR) for the Chinshan plant was submitted to the ROCAEC in December 2010. A review task force was organized by the ROCAEC to perform a very thorough review. As the licensing bases are fully re-examined during the review process, many important issues have been identified and addressed. The key issues resolved include: conformance of SAR to ROCAEC's review guidance; re-examination of post-Fukushima comprehensive safety assessment; qualification of containment protective coatings; GL 96-06 (Assurance of Equipment Operability and Containment Integrity During DBA Conditions); credit for Containment Accident Pressure; issue for Annulus Pressurization Loads Evaluation. These issues required very extensive efforts to resolve. With the cooperative efforts by TPC and contractor (Institute of Nuclear Energy Research), however, all the issues were fully clarified and SAR was approved by ROCAEC on November 15, 2012. The first step SPU (2% OLTP) was successfully implemented in November 2012 at both units. (author)

[en] Chinshan Nuclear Power Plant (NPP) is the first NPP in Taiwan which is a BWR/4 plant. The original rated power for each unit was 1775 MWt. After the project of measurement uncertainty recovery (MUR) for Chinshan NPP, the operating power is 1805 MWt now. The Chinshan NPP Unit 2 cycle 23 stability analyses were performed by the LAPUR6 stability analysis methodology. Comparing the LAPUR6 stability analysis results and vendor's results, they are similar. (author)

[en] Chinshan nuclear power station of Taiwan Power Company (TPC) will implement long term feed water hydrogen injection this year. For the purpose of effective hydrogen water chemistry (HWC), it is necessary and very important to compare plant data and observe the water chemistry change between before and after performing HWC. This paper contains the insoluble and soluble metal characterization, which is based on the recent data before HWC and includes the following working items: 1) The morphology and its compositions of insoluble iron named crud in the inlet of condensate demineralizer. 2) The trend of crud removal efficiency of condensate demineralizer. 3) The compositions of soluble and insoluble metals in the feed water and reactor water. 4) Crud composition prediction deposited on the fuel. Finally, the above results will be explained in detail with respect to the past operational experience as well as the plant-specific water chemistry condition. (author)

No abstract available

[en] In this study the parameter identification tool (PIT) for the loss-of-coolant accident (LOCA) of a BWR-4 plant with Mark-I containment, Chinshan nuclear power plant (NPP) in north Taiwan, is developed. The PIT is composed of the Simplex search algorithm and the MAAP5 code with the integration approach presented by Tsai et al. The LOCA is assumed to occur after station blackout sequence with the break elevation of 17 m and the break area of 2.1*10^-2 m². The break elevation of 16.93 m and the break area of 2.1*10^-2 m² are identified by the PIT with the imitated plant data, which are the RPV pressure and the shroud water level generated by MAAP5 code. With the identified parameters, the timing of events and source terms in the LOCA of Chinshan NPP can be predicted. It demonstrates that the PIT is helpful and important in severe accident management. (authors)

[en] This paper introduces the design evolution, system schemes and design and construction of main nuclear air cleaning components such as HEPA filter, charcoal adsorber and concrete housing etc. for Qinshan 300MW PWR Nuclear Power Plant (QNPP), the first indigenously designed and constructed nuclear power plant in China. The field test results and in-service test results, since the air cleaning systems were put into operation 18 months ago, are presented and evaluated. These results demonstrate that the design and construction of the air cleaning systems and equipment manufacturing for QNPP are successful and the American codes and standards invoked in design, construction and testing of nuclear air cleaning systems for QNPP are applicable in China. The paper explains that the leakage rate of concrete air cleaning housings can also be assured if sealing measures are taken properly and embedded parts are designed carefully in the penetration areas of the housing and that the uniformity of the airflow distribution upstream the HEPA filters can be achieved generally no matter how inlet and outlet ducts of air cleaning unit are arranged

[en] An anticipated transient without scram induced by main steam isolation valve closure (AMSIV) could subject a nuclear power plant to the most severe of accident conditions. The Chinshan plant analyzer contains a complete boiling water reactor system model and can be revised easily for the user's purpose. These features make the Chinshan plant analyzer suitable for AMSIV analysis. The capability of the Chinshan plant analyzer to analyze an AMSIV is illustrated. An AMSIV is simulated, and the simulation results are similar to the results of other research. Furthermore, the AMSIV response of reducing reactor power by decreasing reactor coolant inventory is simulated, and the results of the simulation are similar to those of other research. During this transient, the reactor power is decreased. However, the margin to core uncovery is also decreased. In addition, a method of reducing the reactor power by increasing the feedwater temperature is studied. The mechanism of reducing the reactor power is associated with decreasing the inlet subcooling. Sensitivities of key parameters are also analyzed. A large negative void coefficient causes in undesirable large peak in the reactor power. A small recirculation pump moment of inertia decreases the reactor power

[en] Leakage resulting from sensing line broken at the weldolet to the recirculating riser was found in Chinshan Nuclear Power Plant. A systematic failure root cause analysis, including fractographic examinations, vibration tests, stress analysis, and failure scenario reconstruction, was established. The root cause for this failure has been identified and effective corrective actions can be implemented. (author)

[en] Chinshan is a Mark-I boiling water reactor nuclear power plant (NPP) located in north Taiwan. The severe accident management guidelines (SAMGs) of Chinshan NPP are developed based on the BWR Owners Group (BWROG) EPG/SAG. The Chinshan SAMGs are developed at the end of 2003. MAAP4 code is used as a tool to validate the SAMGs strategies. The development process and characteristics of Chinshan SAMGs are described. Chinshan NPP incorporates several severe accident mitigating features. The 5th emergency diesel generator was added to reduce the core damage frequency (CDF) due to station blackout (SBO) accident. Two raw water tanks in the mountain and firewater truck can inject water into RPV via RHR loop. A T₅U_tX_c sequence, the highest CDF in probabilistic risk assessment (PRA) insight of Chinshan NPP, is cited as a reference case for SAMGs validation. T₅ means failure of long term feedwater system to provide core makeup water. U_t means failure of high pressure coolant injection and reactor core isolation cooling system to provide core make up water. X_c means operator fails to open safety relief valves (SRV) according to the procedure. Low pressure emergency core cooling systems are not able to provide core makeup water because of RPV high pressure. That results all safety injection systems are not operated in the T₅U_tX_c sequence. The severe accident progression is simulated and the entry condition of SAMGs is described. Then the T₅U_tX_c sequence is simulated with RPV depressurization. Mitigation action based on SAMGs of Chinshan NPP are then applied to demonstrate the effectiveness of SAMGs. Sensitivity studies on reactor pressure vessel (RPV) depressurization with the reactor water level and minimum RPV injection flow rate are also investigated in this study. The result shows that RPV depressurization before the reactor water level reaches one fourth of the core water level can prevent the core from damage in the T₅U_tX_c sequence. Flow rate of two control rod drive (CRD) pumps is enough to maintain the reactor water level above TAF and cool down the core in the T₅U_tX_c sequence without operator action. (author)