Chinshan Nuclear Power Plant Research Articles

BWR Parametric Sensitivity Effect of Regional Mode Instability on Stability Boundary

Both in-phase (core wide mode) instability and out-of-phase (regional mode) instability are of great concerns in BWR stability issues. Normally, decay ratios for regional mode oscillations are much less than those under core wide conditions. However, under certain observation mode, the regional mode instability has the phenomenon of power increasing in one half of the core and at the same time, it decrease in the other half, so it looks like that the average power remains essentially constant. This research presents a study of fractional change of decay ratio to evaluate parametric effects of regional mode instability on reload core design power/flow stability boundary for the Chinshan Nuclear Power Plant Unit 2 Cycle 21 (BWR4). Making use of LAPUR5.2 and SIMULATE-3 codes, we have established a methodology to conduct such out-of-phase stability analysis. Many important parameters, such as system pressure, core flow rates, moderator void fraction, fuel physical and geometrical properties, have strong influences on regional mode stability. Current investigations have shown that at some operation points along the stability boundary, certain parameters present more sensitive characteristics.

ICONE15-10260 IN-PHASE AND OUT-OF-PHASE MODES STABILITY ANALYSIS WITH LAPUR5 CODE FOR CHINSHAN

A study of the in-phase (global, core-wide) and out-of-phase (regional) modes stability analysis with LAPUR5 code for Chinshan NPP(nuclear power plant) is presented. The decay ratios are in acceptable agreement with those from vendor's STAIF code for Chinshan NPP Unit 2 Cycle 22 reloads design. Parametric studies for stability analysis have shown that exposure and control rod position are the most important parameters that might affect decay ratios for in-phase and out-of-phase modes stability analysis. The increase in the exposure causes these effects: (a) the increase in the density reactivity coefficient, (b) the change in the axial power shape from bottom-peaked power shape to top-peaked power shape. Additionally, the change in global decay ratio is depended on the change in density reactivity coefficient and axial power shape, and the regional decay ratio is higher than global decay ratio in the BOC condition and bottom-peaked power shape. The more control rods inserted condition causes the increase in the density reactivity coefficient, global and regional decay ratios.

ICONE15-10254 BWR PARAMETRIC SENSITIVITY EFFECT OF REGIONAL MODE INSTABILITY ON STABILITY BOUNDARY

Both in-phase (core wide mode) instability and out-of-phase (regional mode) instability are of great concerns in BWR stability issues. Normally, decay ratios for regional mode oscillations are much less than those under core wide conditions. This research presents a study of fractional change of decay ratio to evaluate parametric effects of regional mode instability on reload core design power/flow stability boundary for the Chinshan Nuclear Power Plant Unit 2 Cycle 21 (BWR4). Making use of LAPUR5 and SIMULATE-3 codes, we have established a methodology to conduct such out-of-phase stability analysis. Many important parameters, such as system pressure, core flow rates, coolant inlet subcoolings, and moderator void fraction, fuel physical and geometrical properties, have strong influences on regional mode stability. Current investigations have shown that at some operation points along the stability boundary, certain parameters are sensitive enough so that the regional mode decay ratio could become greater than the core wide mode decay ratio.

Effectiveness of the Raw-Water System as a Severe-Accident-Management Measure in Station Blackout Accidents

Chinshan is a Mark-I boiling water reactor nuclear power plant (NPP) located in north Taiwan. It incorporates several severe-accident-mitigating features, especially two raw-water tanks in the mountain. According to a probabilistic risk assessment (PRA) of Chinshan NPP, station blackout (SBO) sequences are the most dominant sequences in internal core damage frequency. No credit is taken for the raw-water system in the development of a Chinshan PRA. Therefore, two dominant sequences (T3UTERDGX and T3UTERDG) of the SBO in the Chinshan PRA are cited as reference cases to evaluate the capacity of the raw-water system in the PRA and severe accident. The T3UTERDGX sequence is initiated by loss of off-site power (T3) followed by failure of both diesel generators (DG), failure of gas turbine generators, and failure to recover alternating current (ac) power (ER). That results in loss of all on- and off-site ac power. The high-pressure injection systems fail (UT) initially and timely reactor depressurization fails (X). The T3UTERDG sequence is the same as the T3UTERDGX sequence, except for failure of timely reactor depressurization (X). The MAAP4 code is used as a tool to evaluate the effectiveness of the raw-water system. Based on MAAP4 analysis, the raw-water system cannot cool down the core in the T3UTERDG sequence after introducing severe-accident-management guidelines. The raw-water system cannot flood dry-well water level above minimum debris submerge level (MDSL) in the T3UTERDGX sequence after reactor pressure vessel (RPV) breach. Sensitivity studies show that raw-water injection before the vessel water level reaches level 2 (L-2) can keep core coolability in the T3UTERDG sequence. Three times the raw-water injection rate is the minimum flow rate to flood the dry-well water level above MDSL and cool down the corium on the dry-well floor in the T3UTERDGX sequence. A raw-water system can be used as a mitigating measure, especially in an SBO. The RPV should be depressurized as quickly as possible if a raw-water system is the only mitigation measure in the accident. It is worthwhile to increase the raw-water flow rate to cool down the debris in the dry well after RPV breach.

Reevaluation of the emergency planning zone for nuclear power plants in Taiwan using MACCS2 code

According to government regulations, the emergency planning zone (EPZ) of a nuclear power plant (NPP) must be designated before operation and reevaluated every 5 years. Corresponding emergency response planning (ERP) has to be made in advance to guarantee that all necessary resources are available under accidental releases of radioisotope. In this study, the EPZ for each of the three operating NPPs, Chinshan, Kuosheng, and Maanshan, in Taiwan was reevaluated using the MELCOR Accident Consequence Code System 2 (MACCS2) developed by Sandia National Laboratory. Meteorological data around the nuclear power plant were collected during 2003. The source term data including inventory, sensible heat content, and timing duration, were based on previous PRA information of each plant. The effective dose equivalent and thyroid dose together with the related individual risk and societal risk were calculated. By comparing the results to the protective action guide and related safety criteria, 1.5, 1.5, and 4.5 km were estimated for Chinshan, Kuosheng, and Maanshan NPPs, respectively. We suggest that a radius of 5.0 km is a reasonably conservative value of EPZ for each of the three operating NPPs in Taiwan.

Analysis of Severe Accident Management Strategy for a BWR-4 Nuclear Power Plant

The Chinshan nuclear power plant (NPP) is a Mark-I boiling water reactor (BWR) NPP located in northern Taiwan. The Chinshan NPP severe accident management guidelines (SAMGs) were developed based on the BWR Owners Group Emergency Procedure Guidelines/Severe Accident Guidelines and were developed at the end of 2003. The MAAP4 code has been used as a tool to validate the SAMG strategies. The development process and characteristics of the Chinshan SAMGs are described. The T5UtXC sequence, the highest core damage frequency in the probabilistic risk assessment insight of the Chinshan NPP, is cited as a reference case for SAMG validation. Not all safety injection systems are operated in the T5UtXC sequence. The severe accident progression is simulated, and the entry condition of the SAMGs is described. Then, the T5UtXC sequence is simulated with reactor pressure vessel (RPV) depressurization. Mitigation actions based on the Chinshan NPP SAMGs are then applied to demonstrate the effectiveness of the SAMGs. Sensitivity studies on RPV depressurization with the reactor water level and minimum RPV injection flow rate are also investigated in this study. Based on MAAP4 calculation and the default values of the parameters calculating the severe accident phenomena, 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 T5UtXC sequence. The flow rate of two control rod drive pumps is enough to maintain the reactor water level above the top of active fuel and cool down the core in the T5UtXC sequence without operator action.

Analyses of radioactive aerosols to support accurate internal dose assessments at Chin-Shan nuclear power plant.

Studies of the physical and chemical properties of radioactive aerosols produced in different work environments were made at Chin-Shan nuclear power plant in Taiwan. These properties, including particle mass- and activity-size distributions, elemental and radionuclide compositions, and solubility lung class, are essential for the assessment of internal dose equivalents from the inhalation of radioactive aerosols. Several practices and environments studied include thermal arc cutting, grinding, sandblasting, and noble gas progeny. For physical analyses, cascade impactors, scanning electron microscopy with energy-dispersive x-ray analysis, inductively coupled plasma atomic emission spectrometry, and neutron activation analysis were applied. For chemical analyses, the solubility lung class was determined, in vitro, with the measurement of aerosol dissolution rates in simulated biological fluids. Results of these analyses were used to calculate the annual limit on intake, the derived air concentration, and committed dose equivalents for radiological protection against radioactive aerosols produced at Chin-Shan nuclear power plant.

Thermal-Hydraulic Analysis of a Hypothetical Complete Failure of One Shroud Access Hole in a Boiling Water Reactor

Cracks have been observed in the welds of the shroud support access hole cover plates at Peach Bottom Unit 3 as well as at the Chinshan nuclear power plant. The cracks could potentially result in complete weld failure and, as a consequence, a coreflow bypass. Analysis of the postulated complete weld failure of the Chinshan nuclear power plant shroud support access hole cover plate is performed with RETRAN-02/MOD3. The results show that some confusing phenomena could be observed. The reactor operator should be aware of the phenomena to avoid the likelihood of misoperation. Also, a special procedure needs to be established to deal with this event.