Chinshan Nuclear Power Research Articles

The RADTRAD analysis methodology for the Fuel Handling Accident during the long-term shutdown period of Chinshan Nuclear Power Plant

Abstract This study presents a methodology for the Fuel Handling Accident (FHA) analysis during the long-term shutdown period with the fuels still residing in the reactor core for the Chinshan BWR/4 Nuclear Power Plant. This methodology can be used to calculate the radiation dose received by the control room personnel, assuming failure in safety systems, facilities or structures. The computer code RADTRAD is employed to perform the transport and release of radioactive inventories first, and then calculate the control room dose based on appropriate dose conversion factor for the FHA. The results show that even failure in safety systems, facilities or structures, the whole body dose, thyroid dose and beta-skin dose are well below the acceptance limits specified in the Standard Review Plan.

Development of an Activity Transport Model and in situ Measurements on Deposited Radionuclides in the Primary Coolant Pipelines of the Chinshan Nuclear Power Plant

Activated corrosion products deposited on the surfaces of fuel rods and pipelines contribute the majority of the radiation level in the primary system piping of a light water reactor and would have a significant impact on the safety of maintenance personnel or those involved in future decommissioning work. A computer model for site-specific applications, by the name of ACP_BWR, was developed to predict the distribution of activated corrosion products in the primary coolant circuit of a boiling water reactor (BWR). The prediction results were in reasonably good agreement with the data taken by periodic and in situ measurements at three locations after permanent shutdown of the BWR. Our analyses indicated that the 60Co, 54Mn, 58Co, and 59Fe activities in the core bypass, upper plenum, and lower downcomer regions were higher than those at other regions of the Chinshan Unit 1 reactor. Accordingly, the dose rates resulting from the activated corrosion products deposited at regions close to either side of the core shroud were comparatively high, surpassing those induced by neutron activation at these regions.

International case survey and literature review of maintenance strategy for decommissioning nuclear power plant during transition period

The Chinshan nuclear power plant in Taiwan has currently shutdown when the operating license expired in 2018. However, because the operation license of the independent spent nuclear fuel dry storage facility has not been approved yet, and the fuel storage racks of Spent Fuel Pool (SFP) has almost saturated, there is not enough space to accommodate all irradiated fuels from the reactor core during the transition period. Therefore, part of the mechanical systems, plant structures and electrical equipment of the plant still have to meet the functional requirements for maintaining the safety under the pre-decommissioning activities.In this article, we introduce the international cases of permanent shutdown but the fuel still in the reactor core, and also describe the actions of the Taiwan Power Company in response to this situation. For the providing regulatory advice, we have also proposed a framework of the maintenance strategy associated with important issues for the transition period since 2018, including a revised Aging Management Program (AMP) for keeping the safety of spent nuclear fuel. From the knowledge level and important issues related to the topic, the Phenomena Identification and Ranking Table (PIRT) is quoted to complete the understanding level of requirements during the transition period.

Methodology to incorporate damaged BWR spent fuel as contents in the NAC-UMS® storage system

Following Taiwan’s nuclear waste management policy, INER launched Taiwan’s first Independent Spent Fuel Storage Installation (ISFSI) project in 2005. The technology of dry storage system adopted is NAC-UMS system. INER made some design modifications of NAC-UMS system in order to accommodate the constraints at Chin-Shan Nuclear Power Plant (CSNPP). The modified UMS was named as INER High Performance System (INER-HPS). The corresponding Safety Analysis Report (SAR) of ISFSI was approved by AEC on November of 2008. Through years of dedication, the first dry run test has been accomplished on November 2012. INER-HPS is capable to cover most of the spent fuels dry storage needs in Taiwan with current NAC-UMS License. The current CoC for NAC-UMS includes undamaged BWR fuel with a burn-up of not greater than 45 GWD/MTU. For the purpose of NAC-UMS to incorporate damaged BWR fuel as approved contents, INER and NAC further enter into cooperation to developing and licensing NAC-UMS to include up to four BWR damaged fuel cans and a new BWR damaged fuel basket. This paper summarized the project progress, results of trial fabrication of damaged fuel can, and the methodology to incorporate damaged BWR spent fuel as contents in the NAC-UMS storage system.

Power uprate operation at Chinshan Nuclear Power Station

This process of requesting operation at thermal power levels above the current licensed power level for nuclear power station is referred to as a Power Uprate (PU). There are many nuclear power stations around the world that have implemented power uprate operation for improving their efficiency and economic performance. The Chinshan Nuclear Power Station (CNPS) performed the 1.66% and 2% power uprate for MURPU and SPU, respectively. After successfully achieving the operation of power uprates, the total increase of electrical output for CNPS has amounted to 35.79 MWe. In addition, it also reaches the reduction of CO2 emission about 0.68 million tons per year compare to natural gas combined cycle (NGCC) power plant. Furthermore, the net increase in yearly revenue was compared to the cost of implementing the MURPU and SPU, and the payback time was less than one year. In addition to the calculation of thermal power uncertainty, the power ascension test results, major operating parameters after power uprates, cost benefit analysis, and calculation of CO2 reduction are also presented in this paper. Moreover, the key operating parameters of the turbine cycle measured after power uprate have established a reference base for monitoring station operating performance and provide useful information to turbine cycle design for nuclear power plant.

Code crosswalk of Fukushima-like simulations for Chinshan BWR/4 NPP using MELCOR2.1/SNAP, TRACE/SNAP, PCTRAN and MAAP5.03

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.

Multiple regression approach to predict turbine-generator output for Chinshan nuclear power plant

Abstract The objective of this study is to develop a turbine cycle model using the multiple regression approach to estimate the turbine-generator output for the Chinshan Nuclear Power Plant (NPP). The plant operating data was verified using a linear regression model with a corresponding 95 percnt; confidence interval for the operating data. In this study, the key parameters were selected as inputs for the multiple regression based turbine cycle model. The proposed model was used to estimate the turbine-generator output. The effectiveness of the proposed turbine cycle model was demonstrated by using plant operating data obtained from the Chinshan NPP Unit 2. The results show that this multiple regression based turbine cycle model can be used to accurately estimate the turbine-generator output. In addition, this study also provides an alternative approach with simple and easy features to evaluate the thermal performance for nuclear power plants.

Safety analysis methodology for Chinshan nuclear power plant spent fuel pool under Fukushima-like accident condition

Abstract Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP disaster occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool (SFP), by using TRACE, MELCOR, CFD, and FRAPTRAN codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP SFP. There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP models. And the transient analysis under the SFP cooling system failure condition (Fukushima-like accident) was performed. In addition, the sensitive study of the time point for water spray was also performed. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Finally, the animation model of Chinshan NPP SFP was presented by using the animation function of SNAP with MELCOR analysis results.

The ultimate response guideline simulation and analysis using TRACE, MAAP5, and FRAPTRAN for the Chinshan Nuclear Power Plant

In this research, the TRACE and MAAP5 model of Chinshan BWR/4 nuclear power plant (NPP) has been established for the simulation and analysis of ultimate response guideline (URG). The main actions of URG are the depressurization and low pressure water injection of the reactor and containment venting. This research focuses to assess the usefulness of the URG under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Chinshan NPP TRACE and MAAP5 model. The second step is the URG simulation and analysis under Fukushima-like conditions by using Chinshan NPP TRACE and MAAP5 model. In this step, the base case without URG case was also performed in order to evaluate the URG effectiveness of Chinshan NPP. In order to confirm the mechanical property and integrity of fuel rods, the final step is FRAPTRAN analysis. According to TRACE and MAAP5 analysis results under Fukushima-like conditions, the URG can keep the peak cladding temperature (PCT) below the criteria 1088.7K. It indicates that Chinshan NPP can be controlled in a safe situation. If under Fukushima-like conditions Chinshan NPP does not perform the URG, the water level may drop below the TAF (top of active fuel) which means a fuel melt safety issue may exist.

The establishment of MELCOR/SNAP model of Chinshan nuclear power plant for Ultimate Response Guideline

After Fukushima Daiichi disaster, the safety analysis of severe accidents became one of the safety concerns in Taiwan. The Emergency Operating Procedure (EOP) cannot cope with a multiple system failure situation under a severe accident since it is a “Symptom-basis” procedure. To deal with that, Taiwan Power Company built up a new strategy for Fukushima-like accident called Ultimate Response Guideline (URG). It is a simple strategy with three main conditions: loss of regular motor driven injection system, loss of all AC power and tsunami/earthquake warning. If two of three happen, the operating procedure will change from EOP to URG and start the main works by following the strategy. There are three main works in URG: controlled-depressurization, line up low pressure injection water and prepare containment venting. In this study, MELCOR2.1 was used to calculate the cases of URG and checked the goal of the strategy that prevents the accident or not. There were three steps in this research. First, a model of Chinshan nuclear power plant (NPP) was built. Second, one was the case with URG and the other was not by using the above MELCOR model. The results were compared to MAAP5.0, TRACE and PCTRAN. Finally, some sensitivity studies of depressurization and water injection rate were done.

The Model Establishment and Analysis of TRACE/FRAPTRAN for Chinshan Nuclear Power Plant Spent Fuel Pool

TRACE is developed by U.S. NRC for the nuclear power plants (NPPs) safety analysis. We focus on the establishment and application of TRACE/FRAPTRAN/SNAP models for Chinshan NPP (BWR/4) spent fuel pool in this research. The geometry is 12.17 m × 7.87 m × 11.61 m for the spent fuel pool. In this study, there are three TRACE/SNAP models: one-channel, two-channel, and multi-channel TRACE/SNAP model. Additionally, the cooling system failure of the spent fuel pool was simulated and analyzed by using the above models. According to the analysis results, the peak cladding temperature response was more accurate in the multi-channel TRACE/SNAP model. The results depicted that the uncovered of the fuels occurred at 2.7 day after the cooling system failed. In order to estimate the detailed fuel rods performance, FRAPTRAN code was used in this research. According to the results of FRAPTRAN, the highest cladding temperature located on the node 21 of the fuel rod (the highest node at node 23) and the cladding burst roughly after 3.7 day.

RPV Depressurization Strategy Analysis for Chinshan Nuclear Power Plant by Using MAAP5

When an accident occurs, operators in nuclear power plants (NPPs) must follow emergency operating procedures (EOPs) or severe accident management guidelines (SAMGs). However, EOPs and SAMGs are symptom-based procedures and guidelines to cope with severe transients and accidents. Operators depend on real-time operating parameters of NPPs to perform each action in EOPs or SAMGs. When a beyond-design-basis accident like the Fukushima Daiichi accident of 2011 occurs, EOPs or SAMGs cannot be performed effectively without adequate information. One lesson learned from the Fukushima accident is that such a situation requires advance preparation regarding the key indicators, the water supply, reactor pressure vessel (RPV) depressurization, and containment venting strategies so actions can be performed with limited manpower and time. After the Fukushima accident, Taiwan Power Company established ultimate response guidelines (URGs) and has implemented them in three operating NPPs. An URG is an event-based guideline developed to manage accidents caused by a compound disaster beyond the design basis. The purpose of this study is to find out the differences of RPV depressurization strategies between EOPs and URGs and to discuss the effect of different RPV depressurization strategies on fuel integrity. The plant responses and accident physical phenomena are simulated using MAAP5. The results show that the RPV water level should be maintained as high as possible and the RPV pressure should be controlled sufficiently low at the beginning of RPV emergency depressurization to avoid core uncovery and assure fuel integrity. The URG provides the better RPV depressurization strategy to respond to a beyond-design-basis accident and mitigate an anticipated severe accident consequence as early as possible.

Soft computing technique for developing turbine cycle model of Chinshan Nuclear Power Plant Unit 2

The objective of this study is to develop a turbine cycle model using the adaptive neural-fuzzy inference system (ANFIS) to estimate the turbine-generator output for the Chinshan Nuclear Power Plant (NPP) owned by Taiwan Power Company. The plant operating data was verified using a linear regression model with a corresponding 95% confidence interval for the operating data. In this study, the key parameters were selected as inputs for the neuro-fuzzy based turbine cycle model. After training and validating with key parameters, including main steam to turbine pressure, condenser backpressure, feedwater flow rate, and final feedwater temperature, the proposed model was used to estimate the turbine-generator output. The effectiveness of the proposed ANFIS based turbine cycle model was demonstrated by using plant operating data obtained from the Chinshan NPP Unit 2. The results show that this neuro-fuzzy based turbine cycle model can be used to accurately estimate the turbine-generator output. In addition, a linear multiregression based turbine cycle model was also developed by using the same parameters in order to compare the performance of the ANFIS based turbine cycle model. The results show that the proposed neuro-fuzzy based turbine cycle model is capable of accurately estimating turbine-generator output and providing more reliable results than the multiregression based turbine cycle model, with regard to estimation accuracy and clearly defined trends. The results of this study also provide an alternative approach to evaluating the thermal performance of nuclear power plants.

A Study of the Containment Venting Strategy for the Fukushima Daiichi Accident

The Fukushima Daiichi accident occurred on March 11, 2011. A seismic event and tsunami induced an extended station blackout plus loss of the ultimate heat sink. Three units progressed into a core melt severe accident. The accident occurred in the emergency operation procedure (EOP) domain. However, this situation was already beyond the scope of an EOP. The operator followed the EOP faithfully, and a core melt situation still occurred. An interesting topic is whether it is possible to avoid this type of accident.The purpose of this study is to survey the Fukushima accident progression with respect to the effect of the containment venting strategy for the Chinshan Nuclear Power Plant EOPs. Under the emergency situation, only a small reactor pressure vessel (RPV) injection system was available. This type of accident may be avoided by an early shift from the EOP to the severe accident guideline (SAG), switching from high-pressure injection to low-pressure injection while the reactor core isolation cooling system is available, gradually lowering the RPV pressure, and maximizing the injection flow rate. The plant responses and accident physical phenomena were simulated using MAAP5. The results show that the consequences of an uncovered core and core melt can be avoided by adopting the proper RPV depressurization and containment venting strategy.

Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

The Spent Fuel Pool Analysis of Chinshan Nuclear Power Plant by Using TRACE/FRAPCON-3.4

After Fukushima nuclear power plant (NPP) event, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of the failure of spent fuel pool cooling for Chinshan NPP by TRACE. In this study, by using the above TRACE results, we focused on the application of FRAPCON-3.4 in the spent fuel pool safety analysis of Chinshan NPP. FRAPCON-3.4 can calculate the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. There are two steps considered in this study. The first step is the verification of the FRAPCON-3.4 by using IFA-431 experimental data. The next step is the fuel analysis of Chinshan NPP spent fuel pool by using FRAPCON-3.4 and the TRACE results.

Application of TRACE and CFD in the Spent Fuel Pool of Chinshan Nuclear Power Plant

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. After Fukushima NPP event, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of the spent fuel pool for Chinshan NPP which also assumed the cooling system of the spent fuel pool failed. The geometry of the Chinshan NPP spent fuel pool is 12.17 m × 7.87 m × 11.61 m and the initial condition is 60 ¢J / 1.013 × 105 Pa. In general, the NPP safety analysis is performed by the thermal hydraulic codes. The advanced thermal hydraulic code named TRACE for the NPP safety analysis is developing by U.S. NRC. Therefore, the safety analysis of the spent fuel pool for Chinshan NPP is performed by TRACE. Besides, this safety analysis is also performed by CFD. The analysis result of TRACE and CFD are similar. The results show that the uncovered of the fuels occur in 2.7 days and the metal-water reaction of the fuels occur in 3.5 days after the cooling system failed.

Analyzing the alternative shutdown cooling behaviors for Chinshan Nuclear Power Plant using CFD simulation

The Chinshan Nuclear Power Plant (CSNPP) is a GE-designed BWR4 plant, having two identical units with rated core thermal power of 1804 MWt each unit. Several alternative shutdown cooling methods driven by natural or mixed convection has been proposed by the plant for studying the core cooling capability when the Residual Heat Removal (RHR) systems are not available or the refueling tasks, such as the In Vessel Visual Inspection (IVVI) work etc., is being proceeded. One of the examples is to connect a pipe from the outlet of the new spent fuel heat exchanger to the reactor cavity. The design of the alternatives shall ensure that the maximum core fluid temperature is limited below the boiling temperature of water. In this study, a Computational Fluid Dynamics (CFD) model is developed to analyze the natural convection phenomena during the shutdown period. Through a series of assumption, modeling and meshing processes, a calculation domain with approximate four million meshes including the RPV, reactor cavity and spent fuel pool, have been solved in this study. The analysis results showed that the passive alternative shutdown cooling system could provide sufficient heat removal capability to maintain the maximum core fluid temperature below boiling temperature. The results also indicated that the alternative shutdown cooling system could be served as an appropriate solution for CSNPP when the RHR is inoperable.

The Study of Loss-of-Coolant Accident Parameter Identification for a BWR-4 Plant

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 postulated 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 . Through the developed PIT, the break elevation of 16.93 m and the break area of 2.1×10-2 m2 are identified with the imitated plant data, the RPV pressure and the shroud water level, generated by MAAP5 code. The timing of events and source terms in the LOCA of Chinshan NPP can be predicted with by the MAAP5 code with the identified parameters. It demonstrates that the PIT is helpful and important in severe accident management.

Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 °C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 °C). Additionally, the peak drywell temperature of 155.3 °C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 °C, which is below the pool temperature used for evaluating the net positive suction head of pumps of the RHR system and the Emergency Core Cooling Systems (96.7 °C). The peak containment pressure and temperature are well below the design value (386.1 kPaG and 171.1 °C). Containment integrity of Chinshan Plant can be maintained under the SPU condition.